Radiation shielding materials and containers incorporating same

ABSTRACT

An improved radiation shielding material and storage systems for radioactive materials incorporating the same. The PYRolytic Uranium Compound (“PYRUC”) shielding material is preferably formed by heat and/or pressure treatment of a precursor material comprising microspheres of a uranium compound, such as uranium dioxide or uranium carbide, and a suitable binder. The PYRUC shielding material provides improved radiation shielding, thermal characteristic, cost and ease of use in comparison with other shielding materials. The shielding material can be used to form containment systems, container vessels, shielding structures, and containment storage areas, all of which can be used to house radioactive waste. The preferred shielding system is in the form of a container for storage, transportation, and disposal of radioactive waste. In addition, improved methods for preparing uranium dioxide and uranium carbide microspheres for use in the radiation shielding materials are also provided.

BACKGROUND OF THE INVENTION

This present invention relates generally to radiation shieldingmaterials, radiation shielding containers and methods for preparing thesame. More particularly, the present invention relates to radiationshielding materials incorporating uranium dioxide and/or uranium carbideand containers for radioactive materials incorporating these shieldingmaterials. This invention also relates to methods for preparing uraniumdioxide and uranium carbide microspheres for use in the radiationshielding materials of the present invention.

Storage, transportation, and disposal of radioactive waste, such asspent nuclear fuel (“SNF”), high level waste (“HLW”), mixed waste, andlow level radiation waste is a growing problem in the United States andabroad. In 1995, the Department of Energy (DOE) estimated that thecommercial SNF inventory was about 30,000 metric tonnes initial heavymetal (“MTIHM”) and is expected to exceed 80,000 MTIHM within twodecades. (1 tonnes=1 metric ton=2,205 pounds). Adding DOE's owninventory of SNF and HLW raises the domestic total to nearly 90,000MTIHM.

Unfortunately, it appears that many U.S. commercial nuclear power plantsdo not have sufficient existing storage capacity to accommodate futureSNF discharges. Moreover, much of the DOE's SNF and HLW inventory iscurrently located in unlicensed storage structures. Many of thesestorage structures will have to be upgraded or replaced, and the SNF andHLW relocated. Thus, there is a need for improved radiation shieldingmaterials and radiation shielding containers incorporating theseshielding materials for the storage, transportation, and disposal ofradioactive materials, including, in particular, SNF waste.

Two principal types of storage methods are generally used for SNF: wetand dry. In wet storage, the SNF is typically immersed in a lined,water-filled pool which performs the dual functions of shielding andheat removal with the assistance of and reliance on active systems. Wetstorage of SNF is generally required for a given period of time (about 5years) after the SNF has been discharged from a nuclear reactor.Thereafter, the SNF can be placed into long term dry storage. Drystorage encompasses a wide spectrum of structures that house the fuel ina dry inert gas environment, with an emphasis on passive system designand operation. In dry storage, the radioactive material is typicallydisposed in dry vaults or dry casks. Dry vault installations generallyutilize a concrete building or other concrete structure for radiationshielding. Dry cask storage, on the other hand, utilizes prefabricatedcontainers including an appropriate shielding material. Because dry caskstorage is usually accomplished more quickly and cheaply, it isgenerally preferred over vault storage. Dry cask storage is alsopreferred at sites having an existing infrastructure for receipt,examination, and loading of SNF for economic and scheduling reasons.

The design and manufacture of a suitable container for the dry storageof SNF involves a variety of factors, such as (1) subcriticalityassurance, (2) shielding effectiveness, (3) structural integrity (i.e.,containment), (4) thermal performance, (5) ease of use, (6) cost, and(7) environmental impact. Other factors that may affect the selectionprocess are whether the design has been previously licensed and actuallyused to store SNF, or, if the design has not been licensed, itsperceived ability to meet applicable regulations and standards.

The first factor in designing a storage container is the maintenance ofsubcriticality. In dry storage, the subcriticality design relies oncontrolling the fissile SNF and SNF spacing, and sometimes incorporatesthe use of neutron-absorbing materials. The subcriticality controldesign of dry storage containers is generally acceptable and does nottypically provide any discriminating factors for selecting one designover another.

The second factor in designing a storage container is shieldingeffectiveness. Shielding effectiveness affects both onsite worker andpublic dose rates during the loading and subsequent storage of SNF. Bothneutron and gamma ray shielding must be provided and ensured throughoutthe life of the storage system. Dry storage technology relies on anumber of solid shielding materials, sometimes in combination, to reducegamma and neutron dose rates. The most common solid shielding materialsare different forms of concrete (low-density, high-density, orhydrogenated), metal (ductile cast iron, carbon steel, stainless steel,lead), borated resin, and polyethylene (for neutrons). Often, in orderto function effectively, metal shielding materials must be combined withadditional materials to enhance their neutron absorbing ability.

The third factor in designing a storage container is structuralintegrity (i.e., containment). Structural integrity ensures that theconfinement boundary around the SNF is maintained under all operationaland postulated accident conditions. All SNF storage technologies arerequired to meet the same standards for structural integrity inaccordance with appropriate codes. Therefore, the selection of asuitable storage technology will include consideration of the structuralintegrity of the proposed design.

The fourth factor in container design is thermal performance. With theexception of steel and cast iron, most shielding materials have inherentlimiting temperatures (i.e., a maximum allowable temperature that islower than the fuel cladding temperature limit). Shielding materialthermal limits include both absolute values of temperature and, in thecase of concrete, temperature gradients that create thermal stresses.Adequate decay heat removal is vital to preventing degradation of thefuel cladding barrier to fission product releases.

Dry storage containers rely on a combination of conduction, convection(natural or forced), and radiation heat transfer mechanisms to maintainfuel cladding temperatures below appropriate long term storage limits.In particular, metal casks rely on a totally passive system for heatremoval. The fuel decay heat, in an encapsulating inert gas atmospherecanister, is transferred to the canister's walls by a combination ofradiation and conduction heat transfer. The canister walls, which are incontact with the metal cask wall, transfer this heat by conduction. Atthe outside of the metal cask, the heat is removed by conduction andnatural convention to the environment. Metal cask typically are notsusceptible to thermal limits, since the metals have a highertemperature limit than that of the fuel cladding. However, in thoseembodiments where the metal casks incorporate additional neutronshielding materials their favorable heat-transfer properties may becompromised.

As with metal casks, concrete casks use a passive heat removal system.Concrete casks, however, have an inherent vulnerability, becauseconcrete's thermal conductivity is a factor of 10 to 40 lower than thatof metal. Thus, in order to remove fuel decay heat and stay below boththe fuel cladding and concrete temperature limits, concrete casks mustinclude labyrinthine airflow passages that allow naturalconvection-driven air to enter the cavity enclosing the canister insidethe concrete and then exit through higher elevation passages in theconcrete to the enviromnent. The need for these airflow passagesintroduces the possibility of an accident in which adequate heat removalis reduced or eliminated because of inlets and/or outlets that areblocked by debris, snow, or even nests and hives. As a result, concretecasks require surveillance of their air inlet and outlet flow passages,thereby increasing the associated life-cycle costs and personnelradiation exposures.

The fifth factor in designing a storage container is ease of use, whichis defined as the lack of complexity involved in the operation andmaintenance of SNF. As noted above, the existence of labrynthine airpassages in concrete casks means that additional operation andmaintenance is required. Ease of use, however, is also related to thecomplexity associated with loading, transport, and storage of SNF. Thus,the weight and size of containers are also of particular importance. Forexample, since many existing storage sites are already equipped with acrane in the storage and receiving facility, it is desirable to utilizecontainers with weights that are within the typical crane capacity of 45to 91 tonnes. Metal casks generally cannot be used with such cranes,because the weight of a fully-shielded metal cask loaded with a largenumber of SNF elements can easily exceed the 91 tonnes limit. Thus, eventhough metal casks have desirable heat transfer characteristics, theadditional weight and size associated with metal systems limits theirapplicability.

Additional size and weight limits are imposed when containers aretransported. The U.S. Department of Transportation and state highwayregulations generally limit the gross weight of a waste-carrying roadvehicle to about 80,000 pounds. Since the typical tractor trailer weighsabout 30,000 pounds, the weight of a transportation container and itscontents should not exceed about 50,000 pounds. Heavier weights can betransported by rail, but maximum container widths (diameters) arelimited to approximately 9 feet to allow for adequate clearance betweentracks. U.S. Nuclear Regulatory Commission regulations require that thecontainer provide certain levels of shielding and be capable ofsustaining certain impact stresses without yielding the waste. The endresult of these regulations is that much of the available weight for thetransportation container and its contents must be expended in providingadequate shielding and a shell that can withstand the designated impactstresses. The resulting thickness of the container walls leaves arelatively small amount of space in the container for SNF.

The sixth factor in designing a storage container is cost. Concretecasks are generally the least expensive, with a typical cost of about$350,000 to $550,000, versus $1 million to $1.5 million for their metalcask counterparts.

The seventh factor in designing a storage container is enviromnentalimpact. Over time, environmental mechanisms can degrade storagecontainers, possibly exposing the SNF directly to groundwater or air.Storage containers and shielding materials that minimize degradation arcpreferred for long term storage and disposal.

In summary, metal casks are desirable because they are known to provideeffective heat transfer and structural integrity. Unfortunately, metalcasks are heavier and more expensive than concrete casks. Furthermore,in most SNF applications, metal casks must incorporate separate neutronshields, which may compromise their favorable heat transfer properties.

Thus, there is a significant need for improved, lower weight and higherheat-transfer shielding materials and, also, for containers forhandling, storage, and disposal of radioactive waste that are superiorin performance, size and cost, while providing acceptable structuralstrength, shielding effectiveness, and carrying capacity.

In light of the shortcomings associated with existing dry storagecontainers and the need for long term management of existing inventoriesof SNF, the DOE began to examine alternative means for thetransportation, storage and disposal of such waste. As a result of itsinvestigation, the DOE recommended that the transport and emplacement ofcommercial spent fuel into a DOE waste repository be accomplished usinga class of containers known as the Multi-Purpose Cask (MPC) andMulti-Purpose Unit (MPU). MPC/MPU containers are intended to perform thethree functions of storage, transport, and disposal by directemplacement into a waste repository. The MPC is a thin-shelledcontainer, without shielding, which, once filled, is not intended to beopened. Proposed MPC/MPU designs use metal canisters requiring massivefabrication techniques. As a result, the estimated costs are three tosix times greater than that of concrete cask designs. Furthermore, theMPC containers hold approximately 12% less SNF than that of concretestorage casks. Finally, since the MPC casks do not include shielding,these casks must be outfitted with overpacks consisting of thick-walledsteel and, typically, a separate, neutron-absorbing material to provideshielding.

Meanwhile, the DOE was investigating management options and alternativeuses for large quantities of depleted uranium hexafluoride (“DUF₆”)stored at gas diffusion plants. Among the various disposal optionsconsidered by the DOE was conversion of the uranium hexafluoride touranium metal, which could be machined for use as a radiation shieldingmaterial. However, the high costs of uranium metal production (around$10/kg), combined with the handling, machining, and environmental costsassociated with the use of uranium metal have historically limited itsuse to only a few small applications. In connection with the design ofthe MPC and MPU, for example, the DOE proposed that depleted uraniummetal be used as an axial shield plug in the MPC and as a gammashielding material for the MPU during transport.

Other applications of depleted uranium metal in the fabrication ofstorage containers includes a container made from a composite containinga fibrous mat of interwoven metallic fibers encased within aconcrete-based mixture that can include depleted uranium metal. Anotherproposed application includes a depleted uranium metal core forabsorbing gamma rays and a bismuth coating for preventing chemicalcorrosion and absorbing gamma rays. Alternatively, a sheet of gadoliniummay be positioned between the uranium metal core and the bismuth coatingfor absorbing neutrons. The containers can be formed by casting bismutharound a pre-formed uranium metal container having a gadoliniumsheeting, and allowing the bismuth to cool.

Still another proposed application incorporates a depleted uranium metalwire wound on the inner shell of a cask to create a radiation shield.And yet another proposed application utilizes a composite radiationshield made up of rods of depleted uranium metal. The spaces between therods contain smaller rods and are backfilled with lead or otherhigh-density material. Still other designs utilize pipes of depleteduranium metal, tungsten, or other dense metal, encapsulatingpolyethylene cores, dispersed in rows of concentric bore holes aroundthe periphery of the cask body. None of these existing designs, however,provides a simple, low-cost, low-weight radiation shielding system fortransportation, storage, and disposal of radioactive waste.

Uranium compounds have also been proposed for use as shieldingmaterials. For example, some investigators have proposed that depleteduranium dioxide (DUO₂) pellets be mixed with a cement binder to form amaterial known as DUCRETE, which could be used as a shielding materialin dry storage containers. The DUO₂ pellets replace the gravel aggregatenormally used in concrete. Due to the increased density of DUO₂,however, the thickness of the shielding layer can be reduced. Thus, astorage container made from DUCRETE will have a greatly reduced weightand diameter compared to conventional concrete casks. In a typical cask,for example, the outer shell thickness can be reduced from approximately2.5 feet for concrete to approximately one foot with DUCRETE. As aresult, the cask diameter is reduced by approximately two-thirds, andthe weight is reduced from approximately 123 tonnes to approximately 91tonnes.

Despite these improvements in size and weight, however, DUCRETE caskssystems suffer from disadvantages similar to those experienced withconcrete casks. In particular, since DUCRETE has a low thermalconductivity and low temperature limit, DUCRETE casks must alsoincorporate labrynthine ventilation gaps. Furthermore, it is notexpected that DUCRETE will be able to retain the uranium dioxide pelletsin its cement matrix for a long period of time due to its high porosityof concrete and to the likelihood of water-cement-uranium dioxidereactions at warm temperatures (90-300° C.). DUCRETE may also beincompatible with expected repository requirements. Hence, the use ofDUCRETE in significant quantities for SNF disposal is questionable.

Nuclear fuel manufacturing plants produce small particles of uraniumdioxide and uranium carbide by powdered metallurgical processes. Theseprocesses generally involve production of a powder of the properparticle size and range, which is then pressed into pellets, sintered,and ground to size. Even though powdered processes have shown success,their capacity is limited due to mechanical complexity, particle size,reactivity, and mass transfer limitations. In practice, line capacitiesare limited to approximately 100 tonnes/year, and maximum plant sizes toaround 1,000 tonnes/year.

It has been proposed that aqueous processes be used to generate uraniumdioxide and uranium carbide. Work on aqueous processes, and inparticular on aqueous gelation processes, began in the late 1960's. Bythe mid-1970's pilot-scale facilities for production of uranium oxideand uranium carbide had been constructed. Experimental and pilot plantstudies focused primarily on the use of uranyl nitrate solutions. Forgelation, these uranyl nitrate solutions were dispersed using singlenozzles into columns of chlorinated solvents such as trichloroethylene(TCE) and perchloroethylene. The resulting microspheres were thenprocessed using multiple washing operations with water and ammoniumhydroxide. The resulting microspheres, typically 0.03 mm to 2 mm indiameter, were incorporated into cylindrical pellets. Unfortunately,these aqueous processes had small throughputs and the processing wasmanually intensive. Thus, for planned capacities greater than 100tonnes/yr, these processes were generally inadequate.

It is anticipated that the demand for shielding materials in accordancewith the present invention will require the production of 5,000 to30,000 tonnes/year of uranium dioxide and/or uranium carbide. Thus,there is a need for improved process capable of producing greater than100 tonnes/year, and preferably 5,000-30,000 tonnes/year, of uraniumdioxide and uranium carbide in reasonably-sized plants with inexpensiveequipment. There is a further need for a process for producingmicrospheres of uranium dioxide and uranium carbide over a wide sizerange (30-1,200 microns). There is also a need for an improved gelationprocess for production of uranium dioxide and uranium carbide directlyfrom uranium hexafluoride. Finally, there is a need for an improvedgelation process that avoids the necessity of converting uraniumhexafluoride to uranyl nitrate in order accomplish gelation. The presentinvention addresses these and other needs.

SUMMARY OF THE INVENTION

Briefly, and in general terms, the present invention resides in animproved radiation shielding material and storage systems forradioactive materials incorporating the same. The shielding material ispreferably formed from a PYRolytic Uranium Compound (“PYRUC”) andprovides improved radiation shielding in comparison with other shieldingmaterials. In accordance with the invention, the shielding material canbe used to form containment systems, container vessels, shieldingstructures, and containment storage areas, all of which can be used tohouse radioactive waste. The preferred embodiment of the shieldingsystem is in the form of a container for storage, transportation, anddisposal of radioactive waste.

The precursor for the PYRUC shielding material is preferably a mixtureof a uranium compound and a binding material. In the preferredembodiment, the uranium compound is depleted uranium dioxide (DUO₂) ordepleted uranium carbide (DUC or DUC₂). The uranium compound ispreferably in the form of small particles, and more preferably in theform of pellets or microspheres, which can be coated or uncoated. Thepresent invention incorporates a number of improvements over prior artmethods for producing uranium dioxide and uranium carbide microspheres,whereby 5,000-30,000 tonnes/year of these microspheres can be producedin reasonably-sized plants and with inexpensive equipment. The improvedgelation process of the present invention permits the use of oil in thegel forming column, deliberate carryover of oils to the sintering stepsfor supplying carbon and hydrogen, use of nitrogen as the sinteringcarrier gas, and use of peroxide for gelation of both uranium oxides andcarbides.

In some cases, the precursor material can simply be cured to form aradiation shielding material. However, in preferred embodiments, theparticles are immersed in a matrix of a binding material, so that thebinding material fills the interstitial spaces and also providesadditional neutron shielding. In accordance with the present invention,the binder is advantageously comprised of (1) a carbonaceous material(such as pitch); (2) a high-temperature resin (such as a polyimide); (3)a metal (such as aluminum powder); and/or (4) a metal-oxide (such asalumina). In addition, materials such as hydrogen, boron, gadolinium,hafnium, erbium, and/or indium in their non-radioactive isotopes, can beadded in the mixture in the appropriate chemical form (usually theoxide) to provide additional neutron shielding effectiveness. Theshielding materials are formed by applying sufficient heat to themixture to cause a pyrolytic reaction that forms a solid material.

The present invention also resides in an method for manufacturingstorage containers utilizing PYRUC shielding materials. In accordancewith the invention, the precursor mixture can be poured or extruded intothe container and then pyrolyized to form a solid shield. In aparticularly preferred embodiment, the precursor starting materials arepoured or extruded into a space formed by the inner and outer wall of acontainer and then pyrolized. The manufacturing process provides maximumflexibility in designing shielding shapes. The walls of the containerprovide the shape, structural support, and missile and drop protection,and also function as the secondary confinement barrier for the depleteduranium. The use of PYRUC simplifies shield manufacture and avoids themassive metal forging and machining activities associated with metalcasks.

PYRUC shielding materials in accordance with the present invention offersuperior gamma and neutron radiation shielding with the desirablethermal properties of metal at a much lower thickness, weight, andlife-cycle cost than conventional materials. Furthermore, the PYRUCshielding materials can be optimized for specific circumstances andsource terms. The use of depleted uranium reduces the assay (enrichment)level of the overall package, which provides for criticality mitigation.Furthermore, since PYRUC shielding materials have high thermalconductivities, the need for labyrinthine air passages and dailyinspections is avoided. Similarly, PYRUC materials have higher thermalconductivities and temperature limits than concrete or DUCRETE and,thus, do not limit the design. In particular, the thermal conductivitiesof PYRUC materials exceed DUCRETE values by 25-100%. The temperaturelimits of carbonaceous PYRUC materials exceed 1000° C. and PYRUCmaterials using other binders have temperature limits above 300° C.Moreover, the high thermal conductivity and the high materialtemperature limit of PYRUC eliminate the need for a separate, innercanister for containing SNF. As a result, the PYRUC shielding materialscan be used in SNF containers with direct contact between the shield'sinner annulus and the basket containing the SNF, which further reducessize and weight.

It is believed that PYRUC-shielded SNF containers will cost about$600,000 to $700,000 each, with the PYRUC component accounting for about$200,000 of the cost. The PYRUC container, although having an initialcapital cost slightly greater than the concrete cask, is expected to besignificantly less expensive than the metal cask while having similaradvantages. Lower life-cycle costs are also expected for the PYRUCcontainer as compared with either concrete or DUCRETE containers, sincePYRUC's superior heat transfer properties will preclude the need forfrequent inspection and subsequent maintenance activities. Thus, PYRUCcontainers should be cost-competitive with traditional containers.

PYRUC is also environmentally desirable because it utilizes a wasteproduct from the nuclear industry (depleted uranium) and, in one form, awaste product from the petrochemical industry (carbonaceous bindermaterial) and converts them to environmentally stable forms. The PYRUCshielding material is also environmentally desirable because it is bothmicroencapsulated and macroencapsulated, and has enhanced leachresistance. As a result, the material is potentially stable for geologictime periods. Thus, by virtue of its composition and expected behaviorin a disposal environment, PYRUC is an environmentally friendlymaterial.

Thus, the present invention satisfies the need for a shielding materialhaving combined shielding performance, high temperature resistance, highthermal conductivity, and environmentally desirable characteristics, andfor smaller, lighter containers for storage, transportation, anddisposal of radioactive materials. While the primary applications forPYRUC are containers for SNF and HLW storage, transport, and disposal,PYRUC shielding materials can also be utilized in radiopharmaceuticalcontainers, ion exchange resins, reactor cavity shielding and activatedmaterials (i.e., made radioactive by neutron absorption) among others.

BRIEF DESCRIPTION OF THE FIGURES AND TABLES

The present invention will be more clearly understood for a reading ofthe following detailed description in conjuction with the accompanyingfigures.

FIGURES

FIG. 1 is a cross-sectional view of a container for storage, transport,and disposal of radioactive material which includes a PYRUC shieldingmaterial in accordance with the present invention;

FIG. 2 is a cross-sectional view of the container shown in FIG. 1 alongthe line 2—2 in accordance with the present invention;

FIG. 3 is a flow diagram setting forth the overall process formanufacture of a container incorporating PYRUC shielding materials inaccordance with the present invention;

FIG. 4.1 is an overview in block form of the gelation process forproducing uranium dioxide microspheres in accordance with the presentinvention;

FIG. 4.1a is an overview in block form of the gelation process forproducing uranium carbide microspheres in accordance with the presentinvention;

FIG. 4.2.1 and FIG. 4.2.2 is an overall process flow diagram andmaterial and energy balances for the production of uranium dioxidemicrospheres in accordance with the present invention;

FIG. 4.2a is an overall process flow diagram and material and energybalances for the production of uranium carbide microspheres inaccordance with the present invention;

FIG. 4.3 is a process flow diagram for a depleted uranium hexafluoridereceiving and volatilization station in accordance with the presentinvention;

FIG. 4.4 is a process flow diagram for a the UO₂F₂ production station inaccordance with the present invention;

FIG. 4.5 is a process flow diagram for an uranyl nitrate formationstation in accordance with the present invention;

FIG. 4.6.1 is a process flow diagram for a carbon suspension formationstation utilized in connection with the production of uranium carbidemicrospheres in accordance with the present invention;

FIG. 4.6.2 is a process flow diagram for an uranyl nitrate solutionadjustment station for manufacture of uranium dioxide in accordance withthe present invention;

FIG. 4.6.2 a is process flow diagram for an uranyl nitrate solutionadjustment station for production of uranium carbide in accordance ofthe present invention;

FIG. 4.7 is a process flow diagram for a gel solution preparationstation in accordance with the present invention;

FIG. 4.8 is a process flow diagram for a gel formation station forproduction of 1,200 micron spheres in accordance with the presentinvention;

FIG. 4.9 is a process flow diagram for a gel formation station for 300micron spheres in accordance with the present invention;

FIG. 4.10 is a process flow diagram for an oil purification system inaccordance with the present invention;

FIG. 4.11 is a process flow diagram for a 1,200 micron spheresetting/washing station in accordance with the present invention;

FIG. 4.12 is a process flow diagram for a 300 micron spheresetting/washing station in accordance with the present invention;

FIG. 4.13 is a process flow diagram for a 1,200 micron sphere dryingstation in accordance with the present invention;

FIG. 4.14 is a process flow diagram for a 300 micron sphere dryingstation in accordance with the present invention;

FIG. 4.15 is a process flow diagram for a 1,200 micron sphere conversionand sintering station in accordance with the present invention;

FIG. 4.16 is a process flow diagram for a 300 micron sphere conversionand sintering station in accordance with the present invention;

FIG. 4.17 is a process flow diagram for a calcium nitrate reconstitutionstation in accordance with the present invention;

FIG. 4.18 is a process flow diagram for a ammonium hydroxide solutionpurification station in accordance with the present invention;

FIG. 4.19 is a process flow diagram for a vertical tube furnace gaspurification station in accordance with the present invention;

FIG. 4.20 is a process flow diagram for a ammonium hydroxidereconstitution station in accordance with the present invention;

FIG. 4.21 is a process flow diagram for an urea and HMTA recoverystation in accordance with the present invention;

FIG. 4.22 is a process flow diagram for a cylinder decontaminationstation in accordance with the present invention;

FIG. 4.23 is a process flow diagram for a waste management station inaccordance with the present invention;

FIG. 4.24 is a process flow diagram for an uranium carbide sinteringstation in accordance with the present invention;

FIG. 4.25 is a process flow diagram for an uranium carbide coatingstation in accordance with the present invention; and

FIG. 5 is a process flow diagram for a graphite route for production ofthe uranium carbide microspheres in accordance with the presentinvention; and

FIG. 6 is a process flow diagram for a peroxide gelation process inaccordance with the present invention.

TABLES

Table 1 sets forth the material properties and estimated costs forvarious shielding materials;

Table 2 sets forth the shielding properties for various shieldingmaterials;

Table 4.2 sets forth the overall material and energy balances forproduction of uranium dioxide microspheres in accordance with thepresent invention;

Table 4.2a sets forth the overall material and energy balances for theproduction of dense uranium carbide microspheres in accordance with thepresent invention;

Table 4.3 sets forth the material and energy balances of the depleteduranium hexafluoride receiving and volatilization station for productionof uranium dioxide microspheres or uranium carbide microspheres inaccordance with the present invention;

Table 4.4 sets forth the material and energy balances for the uranylfluoride production station for production of uranium dioxidemicrospheres in accordance with the present invention;

Table 4.4a sets forth the material and energy balances for the uranylfluoride production station for production of uranium carbidemicrospheres in accordance with the present invention;

Table 4.5 sets forth the material and energy balances for the uranylnitrate formation station for production of uranium dioxide microspheresin accordance with the present invention;

Table 4.5a sets forth the material and energy balances for the uranylnitrate formation station for production of uranium carbide microspheresin accordance with the present invention;

Table 4.6.1 sets forth the material and energy balances for the carbonsuspension formation station for production of uranium dioxidemicrospheres in accordance with the present invention;

Table 4.6.2 sets forth the material and energy balances for the uranylnitrate solution adjustment station for production of uranium dioxidemicrospheres in accordance with the present invention;

Table 4.6.2a sets forth the process flow diagram for a uranyl nitratesolution adjustment station for production of uranium carbidemicrospheres in accordance with the present invention;

Table 4.7 sets forth the material and energy balances for the gelsolution preparation station for production of uranium dioxidemicrospheres in accordance with the present invention;

Table 4.7a sets forth the material and energy balances for the gelsolution preparation station for production of uranium carbidemicrospheres in accordance with the present invention;

Table 4.8 sets forth the material and energy balances for the gelformation station for the production of 1,200 micron spheres inaccordance with the present invention;

Table 4.8a sets forth the material and energy balances for the gelformation station for production of 1,200 micron spheres of uraniumcarbide in accordance with the present invention;

Table 4.9 sets forth the material and energy balances for the gelformation station for the production of 300 micron spheres in accordancewith the present invention;

Table 4.9a sets forth the material and energy balances for the gelformation station for production of uranium carbide microspheres inaccordance with the present invention;

Table 4.10 sets forth the material and energy balances for the oilpurification station for production of uranium dioxide microspheres inaccordance with the present invention;

Table 4.10a sets forth the material and energy balances for the oilpurification station for production of uranium carbide microspheres inaccordance with the present invention;

Table 4.11 sets forth the material and energy balances for the 1,200micron sphere setting/washing station for production of uranium dioxidemicrospheres in accordance with the present invention;

Table 4.11a sets forth the material and energy balances for the 1,200micron sphere setting/washing station for production of uranium carbidemicrospheres in accordance with the present invention;

Table 4.12 sets forth the material and energy balances for the 300micron sphere setting/washing station for production of uranium dioxidemicrospheres in accordance with the present invention;

Table 4.12a sets forth the material and energy balances for the 300micron sphere setting/washing station for production of uranium carbidemicrospheres in accordance with the present invention;

Table 4.13 sets forth the material and energy balances for the 1,200micron sphere drying station for production of uranium dioxidemicrospheres in accordance with the present invention;

Table 4.13a sets forth the material and energy balances for the 1,200micron sphere drying station for production of uranium carbidemicrospheres in accordance with the present invention;

Table 4.14 sets forth the material and energy balances for the 300micron sphere drying station for production of uranium dioxidemicrospheres in accordance with the present invention;

Table 4.14a sets forth the material and energy balances for the 300micron sphere drying station for production of uranium carbidemicrospheres in accordance with the present invention;

Table 4.15 sets forth the material and energy balances for the 1,200micron conversion and sintering station for production of uraniumdioxide microspheres in accordance with the present invention;

Table 4.15a sets forth the material and energy balances for the 1,200micron conversion and sintering station for production of uraniumcarbide microspheres in accordance with the present invention;

Table 4.16 sets forth the material and energy balances for the 300micron conversion and sintering station for production of uraniumdioxide microspheres in accordance with the present invention;

Table 4.16a sets forth the material and energy balances for the 300micron conversion and sintering station for production of uraniumcarbide microspheres in accordance with the present invention;

Table 4.17 sets forth the material and energy balances for the calciumnitrate reconstitution station for production of uranium dioxidemicrospheres in accordance with the present invention;

Table 4.17a sets forth the material and energy balances for the calciumnitrate reconstitution station for production of uranium carbidemicrospheres in accordance with the present invention;

Table 4.18 sets forth the material and energy balances for the Ammoniumhydroxide solution purification station for production of uraniumdioxide microspheres in accordance with the present invention;

Table 4.18a sets forth the material and energy balances for the ammoniumhydroxide solution purification station for production of uraniumcarbide microspheres in accordance with the present invention;

Table 4.19 sets forth the material and energy balances for the verticaltube furnace gas purification station for production of uranium dioxidemicrospheres in accordance with the present invention;

Table 4.19a sets forth the material and energy balances for the verticaltube furnace gas purification station for production of uranium carbidemicrospheres in accordance with the present invention;

Table 4.20 sets forth the material and energy balances for the ammoniumhydroxide reconstitution station for production of uranium dioxidemicrospheres in accordance with the present invention;

Table 4.20a sets forth the material and energy balances for the ammoniumhydroxide reconstitution station for production of uranium carbidemicrospheres in accordance with the present invention;

Table 4.21 sets forth the material and energy balances for the urea andhmta recovery station for production of uranium dioxide microspheres inaccordance with the present invention;

Table 4.21a sets forth the material and energy balances for the urea andHMTA recovery station for production of uranium carbide microspheres inaccordance with the present invention;

Table 4.22 sets forth the material and energy balances for the cylinderdecontamination station for production of uranium dioxide microspheresin accordance with the present invention;

Table 4.22a sets forth the material and energy balances for the cylinderdecontamination station for production of uranium carbide microspheresin accordance with the present invention;

Table 4.23 sets forth the material and energy balances for the wastemanagement station for production of uranium dioxide microspheres inaccordance with the present invention;

Table 4.23a sets forth the material and energy balances for the wastemanagement station for production of uranium carbide microspheres inaccordance with the present invention;

Table 4.24 sets forth the material properties and energy balances forthe uranium carbide and sintering station for production of uraniumcarbide microspheres in accordance with the present invention; and

Table 4.25 sets forth the material and energy balances for the uraniumcarbide coating station for production of uranium carbide microspheresin accordance with the present invention.

DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS

With reference now to the exemplary drawings, and particularly to FIGS.1-2, there is shown, in cross-section, a container 10 in accordance withthe present invention. The container includes a lid 12, a base 14 and abody 16 defining a central cavity 18. The container 10 is used to storewaste material, including, in particular, radioactive waste materials,such as SNF. In this regard, a plurality of pressurized water reactor(“PWR”) assemblies housing waste material are fitted inside of a basketassembly 20 disposed within the container 10, as best seen in FIG. 2.The container 10 can have a variety of geometries. In the embodimentshown in FIGS. 1 and 2, the container is cylindrical, having a circularcross-section. Alternatively, the container could have a cross-sectionthat can be square or hexagonal, among other geometries, in order tofacilitate various packing and storing configurations.

The body 16 includes an inner wall 22 a and an outer wall 24 a therebydefining cavity 26 a. A PYRUC shielding material 28 a is disposed withinthe cavity 26 a. The shielding material advantageously absorbs neutronsfrom neutron-emitting waste materials and gamma rays from gamma-emittingwaste materials. As described in detail, below, during manufacture ofthe container, the PYRUC precursor material is prepared and poured orextruded into the cavity between the inner wall and outer wall of thebody and then pyrolized to form a solid radiation shield. Alternatively,the solid radiation shield may be formed by several sequential castings,forming successive axial and radial rings, thereby allowing the shieldto be tailored to a variety of requirements. For example, it may bedesirable to utilize two radial layers of different PYRUC shieldingmaterials, such as a more dense inner layer which will absorb neutronsmore effectively in combination with a less dense outer layer that willabsorb gamma rays.

The inner wall 22 a and outer wall 24 a are formed from forged steelfrom about 0.10 to about 3.00 inches thick, preferably from about 0.5 toabout 1.0 inches thick. The preferred embodiment shown in FIGS. 1 and 2,is an MPU designed to hold twenty-four PWR assemblies. In thisparticular embodiment, the body 16 is 160 inches in height, the diameterof the central cavity 18 formed by the inner wall is 65.8 inches, theouter diameter of the outer wall is 81.8 inches, and the inner wall 22 aand outer wall 24 a of the body 16 define an eight-inch cavity 26 a. Itwill be understood, however, that the thickness of the inner and outerwalls 22 a and 24 a and size of the cavities 18 can vary according tothe strength and shielding requirements of the container 10 and the sizeof the waste to be contained. Forged steel is desirable because it iseconomical, easy to manufacture, and a reasonably good conductor ofheat. Alternatively, other materials such as carbon steel, stainlesssteel, titanium, aluminum, or the like can be used. While stainlesssteel would be generally more expensive, it provides the additionaladvantage of corrosion resistance.

The lid 12 and base 14 are attached to body 16 and each includes aninner wall 22 b and 22 c and an outer wall 24 b and 24 c which define acavity 26 b and 26 c, respectively. In this particular embodiment, bothcavities are about thirteen inches high and incorporate a PYRUCshielding material 28 b and 28 c. The lid and base are constructed fromthe same materials as are used to construct the body.

The container 10 or any of its components, body 12, base 14 and lid 16,can be manufactured with an inner wall 22 and outer walls 24 that arecoated. Coatings can be used, by way of example, to decreasepermeability or to enhance radioactivity absorbing characteristics ofthe container or for corrosion resistance. Typical permeability coatingsinclude glass coatings, epoxy coatings, and inorganic coatings (such asthose containing silica), galvanizing materials (zinc) and zirconia,among others. The coating thickness is typically from about 1.0 to 2,000microns. As best seen in FIG. 1, a liner 30 is located adjacent to theinner wall 22 a. This liner 30 can be a one inch perforated supportplate constructed from materials such as steel, lead, and the like.

Turning now to the details of the basket 20, as shown in FIG. 2, thebasket 20 is dimensioned to hold multiple PWR assemblies. The centralcavity 18 is equipped with a means (not shown), such as a locking pin,which secures the basket in an upright, centralized position. The basketis a removable compartmentalized structure, preferably made of metal,which is designed to hold assemblies of the radioactive material in asegregated manner. In a preferred embodiment, a number of baskets havingdifferent configurations are interchangeable so that both large (24 or21 PWR) and small (12 PWR) assemblies can be accommodated. It is alsodesirable to equip the container 10 with a lifting trunnion 34 attachedto the body 16. This lifting trunnion advantageously facilitateshandling of the container 10.

In use, the base 14 is attached to the container 10 and the container isfilled with SNF by wet or dry methods. After loading, the lid 12 is sealwelded to the body 16 of the container. Alternately, bolt closures withflexitallic, elastomeric, or metallic o-ring/groove sealing (not shown)can be used to seal the lid. If the container was loaded under water,the water is removed via a drain valve (not shown) and the containerdried with warm nitrogen gas by circulation through a top vent (notshown). Subsequently, nitrogen or helium is introduced, and the vent anddrain are welded to the container 10.

In some embodiments, suitable granular material is added to fill thespaces 36 between the basket and the inner wall 22 of the container 10,thereby improving heat transfer and shielding. For storage applications,this granular material includes carbon spheres and sand, particularlycolemonite sand, which includes boron and bound water. For MPC andrelated applications, uranium oxide and uranium carbide could be added,although adjustments may be necessary to account for varying craneweight limits at particular storage or disposal sites.

Referring now to FIG. 3, an overview of the process for preparation ofPYRUC shielding materials is shown. In accordance with the preferredembodiment of the invention, depleted uranium hexafluoride is convertedby an improved gelation process, discussed below, into microspheres of apyrolytic uranium compound, most preferably, into uranium dioxide,uranium monocarbide, and/or uranium dicarbide microspheres (collectively“uranium carbide or “UC”). In some embodiments, at least two sizes ofmicrospheres are utilized to promote higher spatial densities. Also, insome embodiments, the particles are coated with materials such ascarbon, silica, pitch, metal, or the like. During the gelation process,other uranium-containing materials, such as uranium metal and U₃O₈ canbe incorporated and processed to produce microspheres.

Binding materials are sized and classified to match the size of themicrospheres. Two sizes of binding material can be used to maximize thedensity and minimize pore volume of the shielding material. Themicrospheres and binding material are then mixed and homogenized to forma precursor mixture. The precursor mixture is poured or extruded intothe cavity 26 defined by the inner wall 22 and outer wall 24 of thecontainer 10. Heat treatment and pressure are advantageously used topyrolize the microspheres and form a solid shielding material.Inspections and sealing complete the assembly of the container 10.

The precursor mixture contains from about 5 to 100% of a particulatepyrolytic uranium compound. Preferred mixtures contain uranium dioxideand/or uranium carbide. microspheres. The size of the particles can allbe the same size (uniform), can be distributed over a range of sizes(distributed), or can be classified into several discrete size ranges(classified). Preferred particle sizes range from 0.030 mm to 2.0 mm.Smaller particles can be used, but are generally too fine for easyhandling and create environmental concerns. Larger particles can also beused, but require long times for densification, as by sintering, and donot pack as well.

The preferred particle shape is spherical, but particles can be anysuitable shape, including cylindrical, rectangular, and/or irregular.The preferred embodiment uses spherical particles of two discrete sizeranges: 300 to 500 microns and 1,000 to 1,300 microns in diameter,including, in particular, a mixture of 300 micron and 1,200 micronspheres. It is believed that these particles provide a suitablecombination of packing, handling, environmental and densificationrequirements. In a particularly preferred embodiment, the precursormixture contains 80% pyrolytic uranium microspheres. Various binders oradditives make up the remaining portion of the material. Themicrospheres, in turn, are preferably comprised of 70% uraniummonocarbide coated with pyrolytic carbon, as a 1,000 to 1,300 microndiameter particle, and 30% uranium dioxide coated with pyrolytic carbon,as a 300 to 500 micron particle.

As noted above, in preferred embodiments, the binding materials areadded to fill the interstitial spaces, provide additional shielding, andenhance the overall performance of the shielding material. The bindingmaterials generally constitute up to 95% of the precursor mixture.Typically, a binding materials is selected based upon an assessment ofthe radiation spectrum of the material requiring shielding.

The main categories of precursor mixtures in accordance with the presentinvention are classified by the binding material utilized in theirproduction: (1) carbonaceous binders; (2) resin binders; (3) metalbinders; and (4) metal oxide binders. Suitable carbonaceous binders areformed by the low temperature pyrolysis (heating) of pitch, tar,polyvinyl alcohol and related compounds, graphite, coke byproduct or thelike. The preferred form of carbonaceous binder is pitch, because itmixes well with the pyrolytic uranium compound and forms a continuousstructure upon pyrolysis. The carbonaceous binders are preferablypyrolized to the empirical formula C₁H₀₋₂, with C₁H_(0.5) mostpreferred. An advantage of this combination is that it forms anenvironmentally inert shielding material. When pyrolytic uranium dioxideis mixed with a carbonaceous binder, it is preferred that the uraniumdioxide first be coated with, for example, pyrolytic carbon, for bettercarbon-uranium dioxide adhesion.

Resin binders are polymers and include mixtures of polymers, such aspolyethylene, polypropylene, polyurethane, polyimides, and polyamides.Resin binders provide the advantage of excellent neutron shielding,albeit with some heat transfer penalties. The resin binder can be athermoplastic resin, such as polyethylene, polypropylene, orpolyurethane, that can be melted and extruded as a paste or viscousliquid. Advantageously, however, resin binders are comprised ofnon-thermoplastic resin binders, delineated herein as thermoset resins,which do not melt readily, but which bond when the precursor mixture isheated and/or pressed. Examples of such resins includepolytetrafluoroethylene (sold under the tradename TEFLON), polyamides,polyimides, teflon analogues, FEP (fluorinated ethylene-propylene, whichis a copolymer of tetrafluoroethylene and hexafluoropropylene),polyvinylidene fluoride (sold under the tradename KYNAR), and acopolymer of chlorotrifluoroethylene and ethylene (sold under thetradename HALAR), and PFA (perfluoralkoxy), among others. Polyamidesinclude materials such as nylon-6 and nylon-6,6. Polyimides, on theother hand, have a phthalimide structure and are typically formed fromdianhydrides and diamines containing aryl groups. Polyimides generallyhave high strength, stability, and thermal resistance, in some casesgreater than 500° C. Typical polyimides include the reaction products ofbenzophenone tetracarboxylic dianhydride (BTDA) and 4,4′-diaminodiphenylether (DAPE) (sold under the tradenames KAPTON, TORAY, PYRO-ML, andPYALIN), a BDTA/m-phenylenediamine (MPD) derivative (sold under thetradenames MELDIN and SKYBOND), and trimellitic anhydride (TMA)/MPD(sold under the tradenames KERIMID, KERMEL, and ISOMID). In addition, itis believed that both thermoplastic and thermosetting polyfunctionalresins will be advantageously utilized in accordance with the presentinvention. Polyfunctional resins contain at least two chemicalfunctional groups in each repeating polymer unit. In addition topolyurethane and the polyimides and polyamides identified above, othersuitable polyfunctional resins include acetonitrile butadiene styrene(ABS), polyphylene sulfide (PPS), polysulfones, polyesters (includingdacron-type polyesters), phenolic plastics, and fiberglass reinforcedplastic combinations. The preferred resin is both thermosetting andpolyfunctional. In the preferred embodiment, the resin binder is a 100%polyimide resin.

Suitable metal binding materials include copper, zinc, nickel, tin,aluminum, aluminium/boron mixtures and the like. Preferred metal binderscontain aluminum powder. Most preferred is an aluminum/boron mixture,because it exhibits both high heat transfer and neutron shieldingeffectiveness.

Metal oxide binders include both ceramic and refractory materials.Suitable metal oxides include alumina, magnesia, silica, hafnia,hematite, magnetite, silica, and zirconia, among others. Alumina is thegenerally preferred metal-oxide binder. A castable alumina material,with 6% boric/and acid added, is the most preferred, because of itsneutron shielding effectiveness and adhesion to uranium dioxide.

While any one or any combination of the binding materials can be used,the use of one binding material will be preferred for simplicity andgreater mechanical robustness. By way of example, high heat load wasteis advantageously shielded using a shielding material containing abinder having high heat transfer properties, such as a metal binder. Incontrast, mixed uranium-plutonium oxide waste is advantageously shieldedby a shielding material containing a binder optimized for neutronshielding.

The composition of the precursor mixture varies with the category ofbinder material used and application. While the precursor can contain upto 100% of the uranium material (essentially close packing of themicrospheres or pellets), optimum shielding weight is achieved with55-80% pyrolytic uranium compound and 45-20% binder; based on the weightof the precursor mixture.

The precursor mixture also advantageously includes additives, comprisingtypically up to 20% of the binding material, for enhanced shielding,heat transfer, or stability. Typical additives include hydrogen, boron,gadolinium, hafnium, erbium, indium and the like. These additives areincluded in the appropriate chemical forms. For example, an aluminabinder can be combined with boric acid and/or gadolinium oxide. Aparticularly preferred additive is boron-10, which can be added asgranular boric acid and converted to B₂O₃ when the precursor mixture ispyrolized. Alternatively, sodium borate can be utilized. In addition,for gadolinium, halfnium, erbium and indium, the oxide form is generallypreferred. Mechanical additives such as steel shot or glass beads mayalso be added to the PYRUC mixture. Alternatively, additives such asgadolinium, hafnium, erbium, and indium can be added to the gel-formingstep of the gelation process, so that they reside within the spheres ofuranium dioxide/carbide as their respective oxides.

Once the components of the precursor mixture have been selected, theyare combined, and then homogenized. Mixing is advantageouslyaccomplished by either batch or continuous methods, such as twin-screwauger extruders, and slight heating may be applied.

The homogenized mixture is placed within the cavity 26 a formed by theinner wall 22 a and outer wall 24 a of the body 16 by extrusion/pumping(preferred for viscous binder combinations) and/or vibratory methods(preferred for powder blends). Slight heat and pressure may be applied.After filling, sufficient heat (100-1000° C.) and pressure (0-20atmosphere) are applied to the container, to pyrolize and form a solidshielding material. An end closure is attached to the body 16 bysuitable means, such as tungsten inert gas welding in order to seal thebody 16. Thereafter, the container 10 is brushed and polished. Gammaradiography and other non-destructive examination (NDE) methods are usedcheck the body 16 prior to use. The lid 12 and base 14 can be similarlymanufactured.

In those embodiments where a combination of carbonaceous bindingmaterials are employed, the pyrolitic uranium component, carbon powder,additives, and pitch are mixed in an extruder. The extruder thendeposits a first annular layer of the precursor mixture into the cavity26 a, Next the layer is pyrolized in an inert atmosphere of nitrogen,argon or similar gases to form the solid shielding material. Pyrolysistypically requires from about 0.1 to about 24 hours at temperatures offrom about 300-800° C. Thereafter, additional annular layers ofprecursor material are extruded into the cavity 26 a and pyrolized undersimilar conditions.

Alternatively, the inner wall 22 acan be removed for better heattransfer and heat-treated in one step. If carbon powder is used byitself as the binder material, then the mixture is dry-fed into thecavity 18 a. Heat is applied as before and the material is pressed, thusforming the shield. In the preferred embodiment, heat is supplied byelectrical resistance inductance or radiance. Heat may also be suppliedby direct or indirect fired equipment.

For resin-based PYRUC materials, powdered resins are dry blended bymechanical and vibratory means with the uranium form and loaded byvibratory means into the cavity 18 a. Electric heating is preferablyused to heat the material to 400-600° C., typically for 0.1 to 24 hours,to form the PYRUC monolith. If thermal resins are used, they are mixedin an extruder under heat. Thereafter, the mixture is extruded into thecontainer 10 as a viscous fluid. Heat and pressure are then applied toform the solid monolith in a manner similar to the carbon forms.

For metal-based PYRUC materials, the container 10 is heated electricallyor by a fired furnace under an inert cover gas to the melting point ofthe metal binder. For the typical metals cited, this temperature willfall between 400 and 1,000° C., preferably below the melting point ofthe container's materials of construction. Thereafter, an initial amountof molten metal binder is added to the container 10 to form a layer 1 cmto 4 m thick, followed by an initial quantity of preheated uraniummaterial. Due to its density, the uranium material will sink through andto the bottom of the molten metal layer, forming a packed bed of theparticles with the metal filling the interstitial points. The process isrepeated until the cavity 18 a is filled. Thereafter, the heat source isremoved, and the shield cools and solidifies. Alternatively, where apowdered metal, such as a copper or nickel powder, is used, the metalpowder, uranium form, and any additives are dry blended by mechanicaland vibratory means and vibratorily loaded into the cavity 18. Heatingis used to melt the powder, causing the matrix to congeal and fusetogether into a monolith. It is particularly preferred to heat thematerial by induction, utilizing induction coils. As before, typicaltemperatures of 400-1,000° C. and times of 0.1 to 24 hours are required.

For metal-oxide PYRUC materials, the metal oxide and uranium form can becombined with water (0-40 wt %), mixed, and then pumped into the cavity18 a. It is believed that water hydrates the metal oxide binder and,therefore, assists in bonding of the material. The material ispreferably allowed to harden for 2 to 96 hours and then heat treated for0.1 to 24 hours at temperatures up to 400° C. In the preferredembodiment, the shielding material is formed in sequential layers inorder to facilitate heat and mass transfer. Alternatively, the innerwall of the body 16 can be removed and replaced with a temporary,combustible wall (e.g., manufactured from wood products) for casting asbefore. This allows the number of casting steps to be reducedsignificantly, in some cases, allowing a single step. The thermal stepburns away the combustible inner wall.

The choice, mix, and arrangement of the shielding materials used in thePYRUC mixture will vary with the type and quantity of radioactivematerial being transported or stored. Thus, the thickness, diameter,number and arrangement of the shielding materials will be varied toprovide optimum protection against the neutrons and gamma radiationemitted by the particular type and quantity of radioactive material.

The use of uranium dioxide and uranium carbide advantageouslyfacilitates simpler and less expensive manufacturing routes for both theuranium material and the shielding cask. It essentially involves thedirect casting of the PYRUC material from a mixer or an extruder intothe cavity formed between the two metal walls of the cask. Only lowtemperatures are involved, and the casting and machining of uraniummetal are eliminated. Finally, uranium dioxide and coated uraniumcarbide have good heat transfer and thermal characteristics. Thus, theiruse eliminates the need for the labyrinthine air cooling passagespresent in concrete-shielded storage containers, thereby reducingmonitoring requirements and costs.

Table 1 presents a comparison of material properties and estimated costsfor various shielding materials, including PYRUC. As summarized in Table1, PYRUC is a shielding material that provides superior thermalconductivity and temperature limits at a competitive cost, whileoffering superior neutron and gamma shielding.

TABLE 1 Material Properties and Estimated Costs Represen- Represen-Thermal Temper- tative tative Conductivity ature Gamma Neutron Estimated(BTU/ Limit Shielding Shielding Cost Material hr-ft-° F. (° F.)Thickness Thickness ($000) Uranium 15.8 >1000 3.4 18.9 >1,500 MetalPYRUC- 7.8 >1000 7.4 13.3 600-700 UO₂ PYRUC- 16.5 >1000 5.9 14.5 600-700UC Ducrete 5.8 250 14.6 18.5 600-700 Concrete 3.6 250 36.0 36.0 450-550Steel 9.4 >1000 12.0 12.0 1000- 1,500 Cast Iron 29 >1000 12.0 12.0 1000-1,500

Similarly Table 2 presents a comparison of the properties of PYRUC withother shielding materials. As shown in Table 2, casks incorporatingPYRUC typically offer thermal performance and gamma shieldingcapabilities approaching that of metal. Meanwhile, PYRUC materialsprovide low temperature ease of fabrication, and chemically non-reactiveforms which are not susceptible to combustion or chemical interactionabove ground or in an SNF repository.

TABLE 2 Cask Shielding Properties Metal (steel Ideal Cask PropertyConcrete and uranium) Ducrete PYRUC Shield Gamma Fair Excellent GoodExcellent Excellent Shielding Neutron Excellent Fair Very ExcellentExcellent Shielding Good Temperature. 200 1000 (U) 200 >1000 >750 Limit,° F. 2000 (Steel) Clad Limit Thermal Low High Low High High ConductivityIntegrity <300 300.(U) >300 >1000 >1000 Time Frame, 1000. Years (Steel)Cost Low High Medium Medium Low Weight High Medium Medium Medium LowInternal Air Yes No Yes No No Passage Reg. Cask Storage Multi- StorageMulti- Multi- Applicability Purpose Purpose Purpose

Other potential applications for PYRUC include radiopharmaceuticalcontainers, ion exchange resins, reactor cavity shielding, and activatedmaterials. PYRUC may also have utility in other applications as ashielding material for utility resin shields, reactor cavities, navalreactors, spacecraft, and Greater Than Class C (GTCC) materials.

As discussed above, in the preferred embodiment, it is desirable toutilize substantially spherical uranium dioxide or uranium carbideparticles of generally less than 1,300 microns. There are several typesof processes known in the art that can be used to produce suchparticles: (1) Powder Metallurgy Processes (Granulation processes); (2)Melting Processes (Arc glazing, Plasma burner glazing, Suspensionmelting, Glazing of hydrate salts); and (3) Fluid Processes (Syntheticresin condensation, Emulsion processes, and Gelation processes). Inaccordance with the present invention, the small particles of uraniumdioxide and uranium carbide can be generated by any suitable means.

In the past, however, only power metallurgy processes have provided thebasis for commercial production and only for commercial production ofuranium dioxide particles. Processes for the production of uraniumdioxide and uranium carbide are also described in Controlled NuclearChain Reaction: The First 50 Years, American Nuclear Society, 1992, LaGrange Park, Ill., and M. Benedict, T. Pigford, and H. Levi, NuclearChemical Engineering, Second Edition, McGraw-Hill, New York, N.Y., 1981,incorporated herein by reference.

Such processes are mechanically intensive. They typically start with alow density uranium dioxide powder, produced in a rotary kiln fromdepleted hexafluoride, followed by mixing, granulation, pressing apellet, sintering, and pellet-grinding to produce dense uranium dioxideparticles. Furthermore, powder process-based plants are generallymodular, small throughput operations. Thus, scale up to the requirementfor uranium materials contemplated for use in accordance with thepresent invention would necessitate hundreds of process lines.

Other nonfluid methods for the manufacture of uranium dioxide wereinvestigated in the late 1970s and early 1980s, which avoid themechanically intensive, powder processes. These alternative processesare described in S. M. Tiegs, et al., “The Sphere-Cal Process:Fabrication of Fuel Pellets from Gel Microspheres,” ORNL/TM-6906,September, 1979; “Fuels Recycle and Development,” (FRAD) Program Review,Battelle Northwest, Sep. 13-15, 1978, and J. M. Pope, “Spherical UC FuelVia Gel-Precipitation,” American Nuclear Society, Annual Meeting, Miami,Jun. 7-11, 1981, incorporated herein by reference. Additionalinformation is available in “Fuels Refabrication and Development (FRAD)Program Review,” Battelle Pacific Northwest Laboratories, Sep. 13-15,1978; “NPR-MHTGR Fuel Development Program,” Idaho National EngineeringLaboratory (INEL), EGG-NPR-8971, June 1990; and R. H. Perry and C. H.Chilton, Chemical Engineers' Handbook, Fifth Edition, New York, N.Y.,1973, also incorporated herein by reference.

Furthermore, granulation processes are suitable for dense uraniumcarbide particles, but cannot produce dense particles of uranium oxide.Melting processes have the drawback of being expensive and yielding anexcessively-large range of particle sizes.

Accordingly, for economic and capacity reasons, it is preferably mostdesirable to generate the uranium particles using gelation processes. Anoverview of the conversion of depleted uranium hexafluoride intospheric, dense uranium dioxide particles by gelation is presented in A.P. Murray, S. Mirsky, P. Hogroian, and S. Krill, “Gelation Conversion OfDepleted Uranium Hexafluoride Into Dense Uranium Dioxide Microspheres,”Third International Uranium Hexafluoride Conference Proceedings, Nov.28-Dec. 1, 1995, Paducah, Ky., incorporated herein by reference. Theseprocesses include variously as sol-gel, gel-precipitation, internalgelation, external gelation, particle fuel, microsphere, and solutionprecipitation processes. In gelation processes, hydrodynamics is used toform spheres of ammonium diuranate (“ADU”), which are subsequentlycured, dried, and sintered into dense uranium dioxide microspherestypically ranging from 30 to 1,500 microns in size. Furthermore, for aspecific size, a narrow-size distribution can be obtained.

Gelation processes are based on the fact that if a colloidal solution(“sol” or “broth”) of a uranium dioxide precursor (e.g., uranyl nitrate)is dispersed into a fluid with which it is immiscible, or only slightlymiscible, spherical droplets are formed which solidify by gelling(hence, the “gel”). The critical part of the processes occurs when thecolloidal solution is dispersed in the fluid. In order to promotegelling, while maintaining droplet integrity, it is necessary to removethe positive charge on the droplets for greater immiscibility andprecipitation potential. This can be advantageously accomplished byeither (a) extraction of water; (b) extraction of acid; or (c) additionof alkali.

Gelation methods are generally classified as either external or internalgelation routes. In external gelation routes, microspheres of uraniumdioxide or uranium carbide are produced by introducing droplets of auranyl nitrate solution into a column containing ammonia gas. As thedroplets fall through the gas, surface tension effects cause them toform spheres of ADU. Due to size effects upon mass transfer, externalgelation generally requires careful design for production of sphereslarger than about 800 microns. In contrast, internal gelation usesaqueous phase immiscibility in an organic liquid as the basis for sphereformation. The gel formers in internal gelation are typically organicoils or solvents containing ammonia-releasing compounds such as aminesthat release ammonia (e.g. hexamethylenetetramine (“HMTA”)). Due tobetter mass transfer, internal gelation can typically produce larger,more uniform microspheres. Furthermore, since there is better heattransfer between the gel former solution and the droplet, shortercolumns with longer residence times can be used. Thus, in the presentinvention, internal gelation is preferred.

MANUFACTURE OF DENSE URANIUM DIOXIDE AND EXEMPLARY PROCESS

FIG. 4.1 provides an overview of the preferred gelation process forproducing uranium dioxide (UO₂) microspheres according to the presentinvention from depleted uranium hexafluoride (“DUF₆”).

Depleted uranium hexafluoride gas is reacted with steam (“H₂O”) toproduce solid uranyl fluoride (“UO₂F₂”) and gaseous hydrogen fluoride(“HF”). The hydrogen fluoride gas is recovered in the anhydrous form,and the uranyl fluoride solid is collected, quenched, and dissolved inwater. Thereafter, any residual hydrogen fluoride in the uranyl fluoridesolution can optionally be removed by distillation. However, asdiscussed in detail below, the resulting uranyl fluoride solution can beused directly in the gelation process for the production of uraniumdioxide microspheres. The presence of residual hydrogen fluoride doesnot significantly affect the gelation steps, and any residual fluoridehydrogen can be removed from the final uranium dioxide product in thesubsequent steps of aging and washing. Alternatively, the uranylfluoride so produced can be further reacted and converted to uranylnitrate which in turn is used to make uranyl dioxide. As shown in FIG. 4conversion of uranyl fluoride is accomplished by adding calcium nitrate(“Ca(NO₃)₃”) to the aqueous uranyl fluoride solution, therebyprecipitating calcium fluoride (“CaF₂”) and forming aqueous uranylnitrate (“UO₂NO₃)₂”). Prior to gelation, the resulting uranyl nitratesolution is adjusted by evaporation, urea is added, and the solutionchilled.

In the preferred internal gelation routes in accordance with the presentinvention, vibrating nozzles are used to disperse the uranyl nitratesolution into droplets which are then introduced into a vertical columnof an immiscible oil, gel-forming solution. The size of the nozzles andthe vibration frequency determine the droplet sizes, and, thus, themicrosphere sizes. As the uranyl nitrate droplets fall vertically withinthe gel-forming solution, heat transfer between the gel-forming solutionand the droplets causes the uranyl nitrate to form ADU from the ammoniaproduced by the decomposition of hexanethylenesracmine (“HMTA”).Preferably, the gel-forming solution will flow in the opposite directionof the droplets to slow the descent of the droplet and permit additionaltime for the uranyl nitrate solution to form microspheres withsufficient strength to avoid sphere deformation at the bottom of thecolumn. Column heating to 50-100° C. advantageously increases theformation of ADU. The ADU gel spheres are collected at the bottom of thecolumn. Typically, the gel spheres are fragile and require carefulhandling to avoid breakage.

The gel spheres are then aged in an ammonium hydroxide solution. Afteraging, the “green” gel spheres are dried at low temperatures to removewater and excess ammonia. Subsequently, a vertical tube furnace convertsand sinters the microspheres under an inert gas-hydrogen atmosphere.While argon and helium are acceptable inert gases, it is preferable toutilize nitrogen due to its low cost and availability. Furthermore,since nitrogen is slightly reactive, it will advantageously form uraniumnitrides in concentrations up to 10,000 ppm, and more typically in therange of 300 to 1,000 ppm, which will also function as a radiationshielding material.

The final, sintered spheres have individual densities usually exceeding95% of the theoretical density for uranium dioxide. Coarse microspheres(e.g., about 1,000 microns in diameter) might provide spatial densitiesof 65-70% of theoretical, and the addition of a finer microsphere (e.g.,about 300 microns in diameter) might provide spatial densities of 80-85%of theoretical. As a result, two or three size fractions are typicallypreferred in order to achieve spatial densities approaching 90%. Forexample, a 60 wt. % fraction of 1,000 micron spheres, a 20 wt. %fraction of 300 micron spheres, and a 20 wt. % fraction of 30 micronspheres can be used to achieve spatial densities in the 90-95% range.

Table 4.2 and Table 4.2a show the overall material and energy balancesfor production of uranium dioxide microspheres and uranium carbidemicrospheres in accordance with the present invention.

TABLE 4.2 Com- Input, Percent of Nominal Output, Percent of Nominalponent 50% 100% 150% 50% 100% 150% UF₆ 14,000 28,000 42,000 0 0 0 UO₂ 00 0 10,640 21,280 31,920 HF 0 0 0  3,180  6,360  9,540 CaF₂ 0 0 0  3,062 6,123  9,185 HMTA 414 827 1,241 0 0 0 Urea 287 574 861 0 0 0 Ammo-1,400 2,700 4,100 (0) (0) (0) nium Hydrox- ide Argon 77 153 230 0 0 0Nitro- 6,511 13,022 19,533 0 0 0 gen Water 28,000 56,000 84,000 0 0 0CaO 2,200 4,400 6,600 0 0 0 (lime) NaOH 6,000 12,100 18,000 0 0 0 Oil578 1,156 1,734 0 0 0 (Col- umn) Elec- 9,100 18,184 27,300 0 0 0tricity, KW Natural 1E6 1E6 1.5E6 0 0 0 Gas, SCM Diesel 6,000 11,50018,000 0 0 0 Fuel, Liters

TABLE 4.2a Com- Input Output ponent 50% 100% 150% 50% 100% 150% UF₆14,000 28,000 42,000 0 0 0 UC₂ 0 0 0 10,500 20,840 31,500 HF_((nh)) 0 00  3,180  6,360  9,540 CaF₂ 0 0 0  3,062  6,123  9,185 HMTA 414 8271,241 0 0 0 Urea 287 574 861 0 0 0 Ammo- 1,400 2,700 4,100 (0) (0) (0)nium Hydrox- ide Carbon 1,425 2,850 4,275 0 0 0 Pigment Argon 600 1,2001,800 0 0 0 Nitro- 6,511 13,022 19,533 0 0 0 gen Water 28,000 56,00084,000 0 0 0 CaO 2,200 4,400 6,600 0 0 0 (lime) NaOH 9,000 17,600 27,0000 0 0 Meth- 1,850 3,726 5,550 0 0 0 ane Propyl- 2,070 4,141 6,210 0 0 0ene Oil 578 1,156 1,734 0 0 0 (Col- umn) Elec 10,000 20,100 30,000 0 0 0tricity, KW Natural 1E6 1E6 1.5E6 0 0 0 Gas SCM Diesel 6,000 11,50018,000 0 0 0 Fuel, Liters

Each of the following steps are now addressed in greater detail:

A. Uranium Hexafluoride Receiving And Volatilization;

B. Uranyl Fluoride Production By Reaction Of Uranium Hexafluoride AndSteam;

C. Uranyl Fluoride Collection, Quenching And Addition Of Water To FormAn Aqueous Uranyl Fluoride Solution;

D. Uranyl Fluoride Solution Distillation To Adjust Residual HydrogenFluoride Concentration;

E. Uranyl Nitrate Solution Formation By Addition Of Calcium Nitrate AndPrecipitation Of Calcium Fluoride;

F. Uranyl Nitrate Solution Adjustment By The Addition Of Urea AndIncrease In Acidity;

G. Gel Solution Preparation And Addition To Uranyl Nitrate Solution;

H. Gel Sphere Formation By Internal Gelation Techniques;

I. Oil Purification;

J. Gel Sphere Aging By Setting/Washing With Ammonium Hydroxide;

K. Gel Sphere Drying And Liberation Of Ammonia And Water;

L. Gel Sphere Conversion And Sintering;

M. Gel Sphere Collection;

The following additional steps are advantageously undertaken inconnection with the overall process design:

N. Calcium Nitrate Reconstitution;

O. Ammonium Hydroxide Solution Purification;

P. Vertical Tube Furnace Gas Purification;

Q. Ammonium Hydroxide Reconstitution;

R. Urea And HMTA Recovery;

S. Cylinder Decontamination; and

T. Waste Management.

A. Uranium Hexafluoride Receiving and Volatilization

FIG. 4.3 depicts the flow sheet for the Depleted Uranium HexafluorideReceiving and Volatilization Station. Correspondingly, Table 4.3 showsthe material and energy balances for the streams of flow identified bynumerals inside circles in FIG. 4.3 (for example, Stream 1 is identifiedby {circle around (1)}. In a similar manner hereinafter in thisspecification, a table having a specific numerical identification willcorrespond with a figure having the same specific numericalidentification (such as Table 4.3 and FIG. .4.3 or Table 4.4 and FIG.4.4).

TABLE 4.3 Stream Number, Metric Tons per Year (Te/yr) Component 1 2 3 45 6 7 UF₆ (solid) 28,000 0 0 51 0 0 0 UF₆ (gas) 0 28,000 28,000 0 0 0 0Steam 0 0 0 0 1,680 0 0 Water (Con) 0 0 0 0 0 1,680 0 Cylinders (2,240)0 0 (2,240) 0 0 0 Facility 0 0 0 0 0 0 2,880 Waste l/yr Total 28,00028,000 28,000 3,411 1,680 1,680 2,880 l/yr Temp., ° C. 25 60 −70 −25−125 −50 25 Flow, 1,507 −511,000 365,000 1,507 144,000 192 2,880liter/hr l/yr

Depleted uranium hexafluoride is obtained from enrichment plants instandard, 14 tonne cylinders; a typical 14 tonne cylinder will containabout 12.5 tonnes of solid uranium hexafluoride. The extra spaceprovides room for expansion, when the solid uranium hexafluoride isheated. These uranium hexafluoride cylinders 1 are received by truck andrail in the Cylinder Shipping and Receiving Station and stored untilneeded. It is desirable to utilize a storage building having capacityfor enough cylinders for one month's operation (i.e. approximately 300cylinders) and additional capacity for storage of an equivalent numberof empty cylinders 4 while they await shipment to the Cylinder DisposalFacility. Prior to use, the uranium hexafluoride cylinders aretransferred to the Full Cylinder Temporary Storage Station.

A feed of uranium hexafluoride gas is obtained by heating the uraniumhexafluoride cylinders in an autoclave. Heating causes the solid uraniumhexafluoride to sublime, so that the pressurized uranium hexafluoridevapor above the phase can be extracted. Heating of the cylinders isachieved by heating air within the autoclave with steam 5 from a SteamPlant. The oven is heated to a temperature sufficient to causesublimation but, below the liquefaction temperature for uraniumhexafluoride (about 150° F.). The heating rate is preferably selected tomaintain the uranium hexafluoride under subatmospheric pressure. In apreferred embodiment, the oven is heated to about 140° F. in one hour.At this temperature, the uranium hexafluoride temperature will be lower,e.g., about 125° F., and the corresponding cylinder pressure will bebelow atmospheric pressure, e.g. 10 psia. At these pressures, stresseson the cylinders are avoided.

The pressure of the uranium hexafluoride gas 4 is then slightlyincreased (5-10 psig) to near atmospheric conditions using a feedcompressor. This increase in pressure also causes the temperature of theuranium hexafluoride gas to increase to about 212° F.

The uranium hexafluoride feed 4 is then directed sent to the UO₂F₂Production Facility for further processing. In the exemplary process setforth in FIG. 4.3, uranium hexafluoride feed rates of about 28,000tones/year, i.e. 3,200 kg/hr, are employed. At these feed rates,approximately 88 tonnes of uranium hexafluoride can be processed on adaily basis, requiring approximately 7 cylinders of depleted uraniumhexafluoride per day. A typical gas diffusion plant has a gaseous feedstation consisting of three ovens: (1) a first oven heating a fullcylinder of uranium hexafluoride (requiring approximately two hours);(2) a second oven supplying gaseous uranium hexafluoride to theenrichment operations at typical feed rates of 1,500-2,000 kg/hr(requiring approximately four to six hours); and (3) an third ovencooling down (requiring approximately 1-2 hours). Thus, a typical gasdiffusion plant can process about four cylinders per day. Nevertheless,the preferred feed rates of 3,200 kg/hr can be achieved by utilizingfive ovens, sequenced as follows: (1) Oven 1: Heating, two hours; (2)Oven 2: Feeding, first two hours; (3) Oven 3: Feeding, second two hours;(4) Oven 4: Feeding, third two hours; (5) Oven 5: Cooling, 1-2 hours(i.e., hot cylinders have to be cooled prior to moving). It would alsobe desirable to have an additional oven as a spare for use duringmaintenance or in the event of a breakdown. Adding two additional ovens,one as a spare for the heating phase and one as a spare for the feedingoperation would also assist in assuring an uninterrupted supply ofuranium hexafluoride at the design flow rate.

The approach set forth above represents “hot feeding” of the cylinders,with pressures exceeding 25 psig. Where concerns exist regarding thepressure rating of the uranium hexafluoride cylinders, cylinderpressurization can be avoided by “cold feeding” the uranium hexafluorideat temperatures below 147° F. using a withdrawal compressor withsublimation from the solid uranium hexafluoride. Since cold feeding isgenerally restricted to lower feedrates, typically in the 360-450 kg/hr(800-1,000 lb/hr) range per cylinder, it would take approximately 35hours to empty a cylinder containing 12.5 tonnes of uraniumhexafluoride. (The feedrate typically drops to 180-200 kg/hr about 400lb/hr when a withdrawal compressor is not used.) Thus, in order toobtain the desired overall feedrate of 3200 kg/hr at 360 kg/hr per oven,the gelation plant would require simultaneous feeding of about 9cylinders. The following oven sequence would be advantageous in thissituation: (a) Oven 1: Heating, two hours maximum; (b) Oven 2-10:Feeding, probably on a 3-4 hour sequence/changeout schedule; (c) Oven11: Cooling, one-two hours; (d) Oven 12: Spare, for heating/feeding; and(e) Oven 13: Spare, for cooling. Once again, it would be desirable tohave an additional oven as a spare for use during maintenance or in theevent of a breakdown.

In order to minimize or avoid cylinder pressurization, the flow sheetsand mass balances set forth in this exemplary process utilize the 100%baseline case for cold-feeding of uranium hexafluoride using thirteenautoclaves and one spare.

After the uranium hexafluoride is discharged, a “heeling” compressor(not shown) is used to reduce the empty cylinder pressure to less than 1psia and the residual uranium hexafluoride “heel” to 4.5 kg (10 lb).Heel compressor requirements shown in Table 4.3 are estimated as 10% ofthe main compressor. Empty cylinders 4 are sent to Empty CylinderTemporary Storage before shipping to Cylinder Shipping and Receivingand, eventually, forwarded to the Cylinder Disposal Facility.

Condensate 6 from the Autoclave Facility is fed to the Condensate Returnand recycled to the steam plant. Facility waste 7, such as personnelprotective clothing and equipment, is sent to the waste treatmentstation.

B. Uranyl Fluoride Production By Reaction Of Uranium Hexafluoride andSteam

FIG. 4.4 depicts the flow sheet for the UO₂F₂ Production Station,including, inter alia, the reaction of depleted uranium hexafluoride gaswith steam to produce UO₂F₂. The material and energy balances for theexemplary process are shown in Table 4.4.

TABLE 4.4 Section 1 Stream Number, Metric Tons per Year (Te/yr)Component 1 2 3 4 5 6 UF₆ (gas) 28,000 0 0 0 0 0 UO₂F₂ 0 0 0 0 Trace 0(solid) Steam 0 1,430 0 3.472 0 0 Water 0 0 3,472 0 0 0 HF 0 0 0 0 6,3600 Facility 0 0 0 0 0 2,880 l/yr Waste Total 28,000 1,430 3,472 3,4726,360 2,880 l/yr Temp., ° C. −70 125 25 125 200 25 Flow, l/hr 365,000122,000 400 300,000 726 2880 l/yr

TABLE 4.4 Section 2 Stream Number, Metric Tons per Year (Te/yr)Component 7 8 9 10 11 12 UF₆ (gas) 0 0 0 0 0 0 UO₂F₂ 0 0 0 24500 0245,000 (solid) Steam 0 0 0 0 0 0 Water 212,360 17,700 0 0 90,862963,000 HF 0 0 6,360 320 0 3,200 Facility 0 0 0 0 0 0 Waste Total212,360 17,700 6,360 24,820 90,862 1,212,00 0 Temp., ° C. 25 25 25 −10025 25 Flow, l/hr 24,240 2,020 726 1,420 10,400 115,000

TABLE 4.4 Section 3 Stream Number, Metric Tons per Year (Te/yr)Component 13 14 15 16 17 18 UF₆ (gas) 0 0 0 0 0 0 UO₂F₂ 24,500 0 24500245000 0 0 solid Steam 0 0 0 0 0 0 Water 96,350 580 96,350 963,50017,660 17,660 HF 320 320 0 0 0 0 Facility 0 0 0 0 0 0 Waste Total121,200 900 120,850 1,208,50 17,660 17,700 0 Temp, ° C. 25 25 25 25 2545 Flow, l/hr 11,500 103 11,500 115,000 2,016 2,020

TABLE 4.4 Section 4 Stream Number, Metric Tons per Year (Te/yr)Component 19 20 21 22 23 24 UF₆ (gas) 0 0 0 0 0 0 UO₂F₂ 0 0 0 0 0 0(solid) Steam 0 0 0 0 0 0 Water 17,660 177,000 212,360 177,00 87,390 0HF 0 0 0 0 0 6,360 Facility 0 0 0 0 0 0 Waste Total 17,660 177,000212,360 177,000 87,390 Temp., ° C. 45 45 45 45 25 100 Flow, l/hr 2,01620,210 24,240 20,210 10,000 726

TABLE 4.4 Section 5 Stream Number, Metric Tons per Year (Te/yr)Component 25 26 27 28 UF₆ (gas) 0 0 0 0 UO₂F₂ 0 0 0 52 (solid) Steam2,066 636 0 0 Water 0 0 636 8,966 HF 0 0 0 0 Facility 0 0 0 0 WasteTotal 2,066 636 636 9,012 Temp., ° C. 125 125 25 25 Flow, l/hr 237,00073,000 73 1,030

Uranium hexafluoride is reacted with steam according to the followingequation:

UF₆+2H₂O→UO₂F₂+4HF_((anh))+256KW_((t))

In the preferred embodiment, uranium hexafluoride gas 1 from the DUF₆Receiving and Volatilization Station and steam 2 from the steam plant 25are combined in a reactor vessel, such as a pyrolysis reactor or kiln.In the exemplary process shown in FIG. 4.4, the uranium hexafluoride gas1 and steam 2 are introduced concurrently into a pyrolysis reactor atabout 200-300° C. This reaction is exothermic and proceedsspontaneously. At temperatures over about 150° C., no excess steam isrequired and this reaction produces essentially anhydrous hydrogenfluoride. Thus, stoichiometric efficiencies are assumed in the flowsheet shown in FIG. 4.4. In order to prevent runaway temperatures andprovide steam for subsequent use in the Steam Plant, the heat ofreaction is advantageously used to generate steam 4 from deionized waterstream 3.

The gas 5 exiting the reactor vessel consists primarily of anhydroushydrogen fluoride gas with traces of entrained uranyl fluoride powder.The uranyl fluoride powder is removed downstream of the reactor vesselusing cyclone separators and filters. The hydrogen fluoride gas 24 canbe subsequently condensed and collected in an HF Storage Facility. Thestored hydrogen fluoride 9 can be dispensed as a saleable product.

C. Uranyl Fluoride Collection, Quenching and Formation Of An AqueousUranyl Fluoride Solution

FIG. 4.4 depicts the flow sheet for the UO₂F₂ Production Station,including, inter alia, the collection of UO₂F₂ and the subsequentquenching and formation of an aqueous uranyl fluoride solution. Thematerial and energy balances for the exemplary process are shown inTable 4.4.

The uranyl fluoride powder 10 formed in the pyrolysis reactor iscollected and removed by a screw auger device (not shown). Since theuranyl fluoride powder formed in the reactor vessel is discharged atelevated temperatures, the powder 10 is preferably quenched in a waterspray and dissolved in a water solution made up of deionized water 23and wash water 28 from the Cylinder Decontamination Station. Thehydration reaction is represented by the following equation:

UO₂F_(2(s))+6H₂O→UO₂F₂.6H₂O+5kCal_((t))

The heat of hydration is removed by the circulation of cooling water 17through coils (not shown) disposed in the quenching system. Theresulting uranyl fluoride solution 13 is transmitted to the DistillationStation for further processing.

D. Uranyl Fluoride Solution Distillation To Adjust Residual HydrogenFluoride Concentration

FIG. 4.4 depicts the flow sheet for the UO₂F₂ Production Station,including, inter alia, distillation of the uranyl fluoride solution toadjust the residual hydrogen fluoride concentration. The material andenergy balances for the exemplary process are shown in Table 4.4.

A small fraction, approximately 5%, of the anhydrous hydrogen fluorideproduced in the pyrolysis reactor may be entrained with the uraniumfluoride powder and, thus, become hydrated in the Quench Reactor. It isexpected that the hydrated hydrogen fluoride in the quenched uranylfluoride solution 13 can be removed by distillation in the DistillationStation. Heat for the distillation is provided by steam 26 provided fromthe steam plant 25. The distilled product 14 can be returned to thepyrolysis reactor as the azeotrope, HF.2H₂O. The sizing of thedistillation column shown in exemplary process in Table 4.6.2 is basedupon 5% carryover of hydrogen fluoride. Facility waste 6 is sent toWaste Management.

The distilled uranyl fluoride solution 15 is shipped to the UO₂F₂Storage facility where it is available for shipping to the UO₂F₂Precipitation Station.

E. Uranyl Nitrate Solution Formation By Addition Of Calcium Nitrate andPrecipitation Of Calcium Fluoride

FIG. 4.5 depicts the flow sheet for the Uranyl Nitrate FormationStation. The material and energy balances for the exemplary process areshown in Table 4.5.

TABLE 4.5 Section 1 Stream Number, Metric Tons per Year (Te/yr)Component 1 2 3 4 5 6 7 UO₂F₂ 24,500 0 0 0 0 0 0 UO₂(NO₂)₂ 0 0 0 0 0(trace) 0 Water 96,350 6,123 0 (trace) 6,123 6,123 51,496 Calcium 0 0(trace) 0 0 0 12,874 Nitrate CaF₂ 0 0 0 0 0 (trace) 0 Facility 0 0 0 0 00 0 Waste Total 120,850 6,123 (trace) (trace) 6,123 6,123 64,370 Temp.,° C. 25 25 25 25 25 25 25 Flow, liter/hr 11,500 700 (trace) (trace) 700700 6,100

TABLE 4.5 Section 2 Stream Number, Metric Tons per Year (Te/yr)Component 8 9 10 11 12 13 14 UO₂F₂ 0 0 0 0 0 0 0 UO₂(NO₂)₂ 30,929 030,929 0 0 0 0 Water 154,295 6,123 148,172 6,123 0 6,123 51,496 Calcium0 0 0 0 0 0 12,874 Nitrate CaF₂ 6,123 6,123 0 6,123 0 6,123 0 (l) (l)(l) Facility 0 0 0 0 2,880² 0 0 Waste Total 191,347 12,24 179,101 12,2462,880² 12,246 64,370 6 Temp, ° C. 25 25 25 25 25 25 25 Flow, liter/hr18,200 930 17,000 700 2,880² 700 6,100

In the past, gelation methods for producing dense uranium dioxideutilized nitrate solutions generated from uranium oxides as chemicalsubstitutes for nitrate solutions from reprocessing nuclear fuels.Nevertheless, it is believed that uranium fluoride solutions can be useddirectly for gel formation without subsequent processing. The direct useof uranyl fluoride solutions uses concentrations between 0.1 and 40%,and preferably 15-25% in uranyl fluoride. However, since nitratesolutions have previously found favorable application in the past, theexemplary process shown in FIG. 4.5 includes the additional steps forforming the nitrate solution. The material and energy balances relatingto this conversion are contained in Table 4.5.

A calcium nitrate solution 14 from the Calcium Nitrate ReconstitutionStation and, as needed, fresh calcium nitrate powder 3 are combined withdeionized water 4 obtained from the Deionized Water Supply 2 in a mixingvessel. The resulting calcium nitrate solution 7 and uranyl fluoridesolution 1 from the UO₂F₂ Storage facility are introduced into aprecipitator vessel. Tramp and washwater 6 from the Slurry Washer/Dryer,described below, may also be introduced into the Precipitator. Theaqueous calcium nitrate reacts with the uranyl fluoride solution to forma slurry 8 of uranyl nitrate and calcium fluoride according to thefollowing stoichiometric relation:

UO₂F_(2(aq))+Ca(NO₃)_(2(aq))→UO₂(NO₃)_(2(aq))+CaF_(2(s))

The heat of reaction is expected to negligible. The slurry 8 istransferred to a liquid cyclone where the calcium fluoride precipitateis removed. A fraction of the calcium fluoride can be recirculated tofunction as seed crystals in the Precipitator.

The calcium fluoride precipitate 9 is washed and dried in the SlurryWasher/Dryer using deionized water 5 from the Deionized Water Supply.The resulting dry calcium fluoride product 11 is transmitted to a CaF₂Storage Facility. Since the uranium concentration associated with thecalcium fluoride is anticipated to be sufficiently low, the storedcalcium fluoride 13 can be dispensed as a saleable product.

The uranyl nitrate solution 10 is sent to the Uranyl Nitrate SolutionAdjustment Station.

Waste 12 from the Uranyl Nitrate Formation Station is forwarded to WasteTreatment.

F. Uranyl Nitrate Solution Adjustment By The Addition Of Urea,Concentrating The Solution, And Increasing The Acidity

FIG. 4.6.2 depicts the flow sheet for the Uranyl Nitrate AdjustmentStation. The material and energy balances for the exemplary process forproduction of uranium dioxide microspheres and uranium carbidemicrospheres arc shown in Table 4.6.2 and Table 4.6.2a, respectively.

TABLE 4.6.2 Section 1 Stream Number, Metric Tons per Year (Te/yr)Component 1 2 3 4 5 6 7 UO₂(NO₃)₂ 30,929 0 0 23,197 23,197 0 0 U(complexed) 0 0 0 4,671 4,671 0 0 Urea 0 0 5,890 5,890 5,890 0 5,316HNO₃ 0 3,110 0 0 0 0 0 Water 148,172 141,214 23,560 30,469 30,469 13,39810,162 Facility Waste 0 0 0 0 0 0 0 Total 179,101 144,324 29,450 64,22764,227 13,398 15,478 Temp., ° C. 25 60 25 60 0 25 25 Flow, liter/hr.17,000 16,475 3,050 4,073 4,073 1,529 1,472

TABLE 4.6.2 Section 2 Stream Number, Metric Tons per Year (Te/yr)Component 8 9 10 11 12 13 UO₂(NO₃)₂ 0 23,197 0 23,197 0 0 U (complexed)0 4,671 0 4,671 0 0 Urea 574 5,890 0 5,890 0 0 HNO₃ 0 0 0 0 0 0 Water 030,469 0 30,469 71,400 71,40 0 Facility Waste 0 0 2,880³ 0 0 0 Total 57464,227 2,880³ 64,277 71.400 71.40 0 Temp, ° C. 25 0 25 38 25 45 Flow,liter/hr. 33 4,073 2,880³ 4,073 8,151 8,151

TABLE 4.6.2a Section 1 Stream Number, Metric Tons per Year (Te/yr)Component 1 2 3 4 5 6 7 UO₂(NO₃)₂ 30,929 0 0 23,197 23,197 0 0 U(complexed) 0 0 0 4,671 4,671 0 0 Urea 0 0 5,890 5,890 5,890 0 5,316HNO₃ 0 3,110 0 0 0 0 0 Carbon 0 0 0 2,850 2,850 0 0 Water 148,172152,614 23,560 30,469 30,469 13,398 10,162 Facility Waste 0 0 0 0 0 0 0Total 179,101 155,724 29,450 64,227 64,227 13,398 15,478 Temp., ° C. 2560 25 60 0 25 25 Flow, liter/hr 17,000 17,727 3,050 4,073 4,073 1,5291,472

TABLE 4.6.2a Section 2 Stream Number, Metric Tons per Year (Te/yr)Component 8 9 10 11 12 13 14 UO₂(NO₃)₂ 0 23,197 0 23,197 0 0 0 U(complexed) 0 4,671 0 4,671 0 0 0 Urea 574 5,890 0 5,890 0 0 0 HNO₃ 0 00 0 0 0 0 Carbon 0 2,850 0 0 0 0 2,850 Water 0 30,469 0 30,469 71,40071,400 11,400 Facility Waste 0 0 2,880⁽³⁾ 0 0 0 0 Total 574 67,0772,880⁽³⁾ 64,277 71,400 71,400 14,250 Temp., ° C. 25 0 25 38 25 45 50Flow, liter/hr 33 4,254 2,880⁽³⁾ 4,073 8,151 8,151 1,356

It has been found that more desirable gelation occurs if the uranylnitrate solution is adjusted prior to gelation. In particular, it isdesirable to decrease the acidity of uranyl nitrate solution and to addurea (CO(NH₂)₂) in order to stabilize the uranyl ion.

Recycled urea 7 from the Urea Recycle Station, described below, iscombined with deionized water 6 obtained from the Deionized Water Supplyin a mixing vessel. The urea is dissolved and, as necessary, urea powder8 is added to form a solution having a molar ratio of 1 to 1.5, andpreferably 1.25, urea to uranium.

The urea solution 3 is added to the uranyl nitrate solution 1 from theUranyl Nitrate Formation Station in a Vapor Recompression (VR)Evaporator. The VR Evaporator provides the benefits of multistageevaporation in a single-stage unit and achieves typical evaporationefficiencies of 0.0452 KW-hr per kg of water (35 BTU/lb of waterevaporated, as compared to normal values of around 1,000 BTU/lb). Theevaporator advantageously performs three functions: (1) mixing the ureawith the uranium solution to form the urea/uranium complex; (2)concentrating the uranium solution; and (3) rendering the solutionslightly acid deficient (i.e., having an anion (nitrate and fluoride) touranium molar ratio of 1.5 instead of 2). The uranyl nitrate solutiongenerated by the VR Evaporator contains uranium in the 4.8-3.0 molarconcentration range (Table 4.6.2 utilizes a value of 2.2 molar) and isdense, with a specific gravity of approximately 4.8. The overheadproduct evolved from the VR Evaporator is a dilute nitric acid solution2 (approximately 2%), which is transferred to the Calcium NitrateReconstitution Station, described below.

As the uranyl nitrate solution is generated in the VR Evaporator it istransferred to a Uranyl Nitrate Storage Tank. The uranyl nitratesolution 4 is then chilled to approximately 0° C. before the chilledsolution 9 is transferred to the Gel Solution Preparation Station. Sincethe solution boiling point elevation data for this solution is notreadily available, the evaporation energy shown in the material andenergy balance in Table 4.6.2 is estimated at 0.1292 KW-hr/kg (100BTU/lb).

Facility waste 10 from the Uranyl Nitrate Adjustment Station is sent toWaste Treatment.

In the preferred gelation process, the uranium feed material is obtainedin the form of uranium hexafluoride. However, Uranium and, inparticular, depleted uranium is available in a variety of forms, such asuranium metal, low density uranium oxides, and uranium tetrafluoride.Thus, it is desirable to incorporate theses forms into the gelationprocess. In accordance with the present invention, the uranyl fluoridesolution beneficially assists the dissolution of alternative feedmaterials. In particular, the uranyl fluoride solution can be used todissolve alternative uranium feed materials. It is believed that thisalternative route will be particularly advantageous because it utilizesfewer reagents, requires less precipitant, and generates less waste.Thus, it is expected that 80-100% of its total uranium feed will beobtained as uranyl fluoride, derived from the uranium hexafluoride, withthe remaining 0-20% of the uranium feed obtained from alternativesources of uranium. It may be desirable to facilitate the process byadding less than stoichiometric amounts of nitric or hydrofluoric acidand 0.0001-0.5% a catalyst such as fluorboric acid (HBF₄), fluorboricacid as a catalyst. Urea at a 1-1.5 molar ratio (to the total uranium)is also added prior to dissolution of the alternative uranium form.After dissolution, 0.01-10% of aluminium, as the fluoride or nitrate, isadded to complex the fluoride ion.

G. Preparation Of Gel-Former Solution Having An Ammonia ReleasingCompound

FIG. 4.7 depicts the flow sheet for the Gel Former Preparation Station.The material and energy balances for the exemplary process forproduction of uranium dioxide microspheres and uranium carbidemicrospheres are shown in Table 4.7 and Table 4.7a, respectively.

TABLE 4.7 Section 1 Stream Number, Metric Tons per Year (Te/yr)Component 1 2 3 4 5 6 7 UO₂(NO₃)₂ 23,197 0 0 0 0 0 0 U (complexed) 4,6710 0 0 0 0 0 Urea 5,890 0 280 0 0 0 0 HMTA 0 0 5,715 827 16,542(1)165,542 16,542 Water 30,469 trace 22,844 0 22,844 228,440 22,844Facility Waste 0 0 0 0 0 0 0 Total 64,227 trace 8,559 827 393,386 39,38639,386 Temp., ° C. 0 25 25 25 25 0 0 Flow, liter/hr. 4,073 trace 3,65050 3,747 37,470 3,747

TABLE 4.7 Section 2 Component/ Stream Number, Metric Tons per Year(Te/yr) Item 8 (3) 9 (3) 10 (3) 11 UO₂(NO₃)₂ 23,197 16,238 6,959 (trace)U (complexed) 4,671 3,270 1,401 (trace) Urea 5,890 4,123 1,767 0 HMTA16,542 11,579 4,963 0 Water 53,313 37,319 15,994 0 Facility Waste 0 0 02,880⁵ Total 103,613 72,529 31,084 2,880⁵ Temp., ° C. 0 0 0 25 Flow,liter/hr. 6,571 (2) 4,600 1,971 2,880⁵

TABLE 4.7a Section 1 Stream Number, Metric Tons per Year (Te/yr)Component 1 2 3 4 5 6 7 UO₂(NO₃)2 23,197 0 0 0 0 0 0 U (complexed) 4,6710 0 0 0 0 0 Urea 5,890 0 280 0 0 0 0 HMTA 0 0 15,715 827 16,542(1)165,542 16,542 Carbon 2,850 0 0 0 0 0 0 Water 30,469 trace 22,844 022,844 228,440 22,844 Facility Waste 0 0 0 0 0 0 0 Total 67,077 trace38,559 827 39,386 393,386 39,386 Temp., ° C. 0 25 25 25 25 0 0 Flow,liter/hr. 4,254 trace 3,650 50 3,747 37,470 3,747

TABLE 4.7a Section 2 Stream Number, Metric Tons per Year (Te/yr)Component 8 9 10 11 UO₂(NO₃)₂ 23,197 16,238 6,959 (trace) U (complexed)4,671 3,270 1,401 (trace) Urea 5,890 4,123 1,767 0 HMTA 16,542 11,5794,963 0 Carbon 2,850 1,995 855 0 Water 53,313 37,319 15,994 0 FacilityWaste 0 0 0 2,880⁵ Total 106,463 74,524 31,939 2,880⁵ Temp., ° C. 0 0 025 Flow, liter/hr. 6,571 (2) 4,600 1,971 2,880⁵

As discussed above, in the preferred embodiment, gelation is preferablyaccomplished using an internal gelation technique. Internal gelation ofthe chilled uranium nitrate solution from the Uranyl Nitrate AdjustmentStation preferably utilizes a gel formation solution comprising anammonia releasing compound, such as hexamethylenetetramine (“HMTA”),(CH₂)₆N₄. Other amines can also be used, such as ethylene diamine(“EDA”).

HMTA powder 4 is dissolved in deionized water 2 from the Deionized WaterSupply in a Mixing Tank. Recycled HMTA solution 3 from the HMTARecycling Station, described below, is introduced to form a solutionhaving about a 3 molar concentration. The HMTA solution 5 issubsequently chilled to around 0° C. The chilled HMTA solution 7 iscombined with the chilled urea-uranyl nitrate solution from the UranylNitrate Adjustment Station in a Static Mixer. The resultingHMTA-containing urea-uranyl nitrate broth is chilled to about 0° C. inorder to avoid HMTA decomposition and premature precipitation of ADU.Even at these reduced temperatures, however, the solution has a limitedshelf-life, on the order of 1-3 days. Stabilizers, such as surfactantsand aliphatic hydrocarbons can be added to the solution to extend itsshelf-life. For example, a low concentration (0.001 to 1%) of asurfactant can be added to the solution to aid sphere formation and toinhibit particle agglomeration during gel formation.

In the exemplary process shown in FIG. 4.7 and Table 4.7, a firstportion 9 containing approximately 70% of the chilled broth solution 8is transferred to a 1,200 Micron Gel Formation Station, and a secondportion 10 containing approximately 30% of the chilled broth solution 8is transferred to a 300 Micron Gel Formation Station. Both of thesefacilities are described in additional detail below. As describedpreviously, a minimum of two particle sizes are desirable; one sizerelatively coarse (1000 to 2000 microns, and preferably 1000 to 1300microns diameter) and one size relatively fine (30 to 1000 microns,preferably 300 to 500 microns in diameter). This allows for closerpacking and higher densities, resulting in better shielding.

Facility waste 11 from the Gel Former Preparation Station is sent toWaste Treatment.

H. Gel Sphere Formation Of Uranium/Ammonium Diuranate Precipitate ByInternal Gelation Techniques

FIG. 4.8 depicts the flow sheet for the 1200 Micron Gel FormationStation. The material and energy balances for the exemplary process forproduction of uranium dioxide microspheres and uranium carbidemicrospheres are shown in Table 4.8 and Table 4.8a, respectively.

TABLE 4.8 Section 1 Stream Number, Metric Tons per Year (Te/yr)Component 1 2 3 4 5 6 7 UO₂(NO₃)₂ 16,238 0 0 0 16,238 0 0 U (complexed)3,270 0 0 0 3,270 0 0 Urea 4,123 0 0 0 4,123 0 0 HMTA 11,579 0 0 011,579 0 0 Water 37,319 0 8,740 <5,000 3,869 33,450 −35,000 Oil 0 0 03.35E7 15,706 6.69E7 6.69E7 Nitrogen 0 4,462 0 0 0 0 0 Facility Waste 00 0 0 0 0 0 Total 72,529 4,462 8,740 3.35E7 54,785 6.69E7 6.69E7 3Temp.,° C. 0 25 150 70 70 70 69 (Steam) Flow, liter/hr. 4,600 409,0001E6 4.8E6 3,474 9.54E6 9.54E6

TABLE 4.8 Section 2 Stream Number, Metric Tons per Year (Te/yr)Component 8 9 10 11 12 13 UO₂(NO₃)₂ 0 0 0 0 0 0 U (complexed) 0 0 0 0 00 Urea 0 0 0 0 0 0 HMTA 0 0 0 0 0 0 Water 66,900 0 33,450 33,450 8,740 0Oil 6.69E7 10 3.35E7 3.35E7 0 0 Nitrogen 0 4,462 0 0 0 0 Facility Waste0 0 0 0 0 2,880⁽⁵⁾ Total 6.7E7 4,472 3.35E7 3.35E7 8,740 2,880⁽⁵⁾ 3Temp., ° C. 69 25 69 69 150 25 Flow, liter/hr 9.54E6 409,000 4.8E6 4.8E61,000 2,880⁽⁵⁾

TABLE 4.8a Section 1 Stream Number, Metric Tons per Year (Te/yr)Component 1 2 3 4 5 6 7 UO₂(NO₃)₂ 16,238 0 0 0 16,238 0 0 U (complexed)3,270 0 0 0 3,270 0 0 Urea 4,123 0 0 0 4,123 0 0 HMTA 11,579 0 0 011,579 0 0 Carbon 1,995 0 0 0 1,995 0 0 Water 37,319 0 8,740 <5,0003,869 33,450 −35,000 Oil 0 0 0 3.35E7 15,706 6.69E7 6.69E7 Nitrogen 04,462 0 0 0 0 0 Facility Waste 0 0 0 0 0 0 0 Total 74,524 4,462 8,7403.35E7 56,780 6.69E7 6.69E7 Temp., ° C. 0 25 150 70 70 70 69 (steam)Flow, liter/hr 4,600 409,000 1E6 4.8E6 3,474 9.54E69 9.54E6

TABLE 4.8a Section 2 Stream Number, Metric Tons per Year (Te/yr)Component 8 9 10 11 12 13 UO₂(NO₃)₂ 0 0 0 0 0 0 U (complexed) 0 0 0 0 00 Urea 0 0 0 0 0 0 HMTA 0 0 0 0 0 0 Carbon 0 0 0 0 0 0 Water 66,900 033,450 33,450 8,740 0 Oil 6.69E7 10 3.35E7 3.35E7 0 0 Nitrogen 0 4,462 00 0 0 Facility Waste 0 0 0 0 0 2,880⁽⁵⁾ Total 6.7E7 4,472 3.35E7 3.35E78,740 2,880⁽⁵⁾ Temp.,° C. 69 25 69 69 150 25 Flow, liter/hr 9.54E6409,000 4.8E6 4.8E6 1,000 2,880⁽⁵⁾

In the Gel Formation Station, spheres of uranium dioxide are preferablyformed by an internal gelation process. In this process, small drops ofthe chilled broth solution 1 from the Gel Solution Preparation Stationare dispersed using vibrating feed nozzles (not shown) into a GelForming Column containing oil. In the past, gelation was accomplishedusing columns of chlorinated solvents such as trichloroethylene (TCE)and perchloroethylene. The preferred embodiment, however, uses an oil,such as kerosene or fuel oil nos. 1, 2 or 3, which are relativelynon-toxic and nonflammable. These oils also have low ash and residuecontents and gum less than other oils, such as heat transfer oils.Further, use of oils eliminates the production of acid gases that isassociated with the use of halogen-containing solvents. Still further,when oil is used, washing operations can be advantageously scaled back,because any oil carried over to the sintering step can be burnt away.Alternatively, oil that is carried over to the sintering process can bepyrolized to carbon for carbide production, dissociated into a coating,or reformed into hydrogen gas.

While the prior art uses high velocity or air impact/impingementnozzles. The preferred embodiment uses low kinetic energy nozzles, suchas ultrasonic nozzles, which reduce energy consumption, reduce gashandling, reduce deformation, increase homogeneity, and provide bettercontrol with wider operational ranges. The vibrating nozzles fragmentthe chilled broth solution into droplets 0.1-6 mm diameter in an airspace above the Gel Forming Columns. Fragmentation of the chilled brothsolution can optionally be undertaken using a nitrogen purge 2. Nozzleflowrates will vary, but are relatively small, on the order of liters ofsolution per hour. As the immiscible droplets fall through the oil,surface tension effects form each droplet into a sphere, the diameter ofwhich is determined by the size and vibration frequency of the feednozzles. The higher temperature of the oil, ranging from 50 to 100° C.,initiates the dissociation of the ammonia-releasing compound (HMTA) andthe formation of a uranium/ADU-like precipitate. Typically, the “green”gel spheres formed in the Gel Formation Station are about three timesthe desired diameter of the final product. Therefore, for a finaluranium dioxide microsphere diameter of 1,200 microns, the green gelsphere diameter should be about 3,600 microns.

The typical minimum free-fall residence time for gel sphere formation inthe Gel Forming Column is in the 20-30 second range, after which the gelspheres are sufficiently well-formed to avoid sticking and deformation.This residence time is achieved by having a column of sufficient height,by countercurrent flow of the oil, or by a combination of both. Thepreferred embodiment uses both.

In the exemplary process described in Table 4.8, the following systemcharacteristics are utilized: (1) nozzle flow rate: 4 liters/min; (2)green sphere diameter: 3,600 microns (0.36 cm); (3) column residencefree-fall time: one minute. Column height is predicated upon one-minuteresidence time, which translates into an active height of approximately22 meters without any credit for the oil flow. Using an upwards oil flowvelocity of 30 cm/sec (1 ft/sec) produces a column height of 4 meters(13 feet). For piping and plumbing connections, flow disengagementsections, and sphere aging, Table 4.8 utilizes a 6.1 meter height (20feet). A four-inch diameter represents the full-length column diameterrequired per nozzle (i.e., no converging/diverging sections). Settingresidence time is sixty minutes. Thus, a nominal column diameter of 30cm (1 foot) would accommodate approximately seven nozzles, have anaqueous feed rate of approximately 28 liters per hour (0.12 gpm), and anoil flow rate of 79,500 liters per hour (350 gpm). In contrast, anominal column diameter of 51 cm (20 inches) would accommodate nineteennozzles, have an aqueous feed rate of 76 liters per hour (0.33 gpm), andan oil flow rate of 223,000 liters per hour (980 gpm). These analysesassume a nominal column diameter of 31 cm (i.e., 7 nozzles) forproducing the 1,200 micron gel spheres.

After the green gel spheres have been formed and fallen to the bottom ofthe column, they will have developed sufficient strength to resistdeformation under their own weight. These green gel spheres remain inthe hot oil for thirty to sixty minutes for setting. Settingadvantageously permits additional precipitation to occur and hardens thespheres. Even after setting, however, only about 1% of the HMTA willtypically decompose, and, thus, only about 5% of the precipitationreaction will have gone to completion. As much as about 30 cm of astatic bed of green spheres will accumulate at the bottom of the GelForming Column during the setting process. These green gel spheres 5 areremoved from the Gel Forming Column and transferred to the 1,200 MicronSphere Aging/Washing Station.

Oil from the column overflows to an Oil Overflow Tank. a portion 7 ofthe of the oil 8 from the Oil Overflow Tank is recirculated through aheat exchanger and the heated oil 6 is returned to the Gel FormingColumn. Since the oil in the Gel Forming Column absorbs some water fromthe gel spheres (up to its saturation limit), a fraction of the oil 10is sent to the Oil Purification Station, described below, for waterremoval and the purified oil 4 returned to the Gel Forming Column.

Facility waste 13 from the 1,200 Micron Gel Formation Station is sent toWaste Treatment.

FIG. 4.9 depicts the flow sheet for the 300 Micron Gel FormationStation. The material and energy balances for the exemplary process forproduction of uranium dioxide microspheres and uranium carbidemicrospheres are shown in Table 4.9 and Table 4.9a, respectively.

TABLE 4.9 Section 1 Stream Number, Metric Tons per Year (Te/yr)Component 1 2 3 4 5 6 7 UO₂(NO₃)₂ 6,959 0 0 0 6,959 0 0 U (complexed)1,401 0 0 0 1,401 0 0 Urea 1,767 0 0 0 1,767 0 0 HMTA 4,963 0 0 0 4,9630 0 Water 15,994 0 4,034 <1,000 9,944 6,050 ˜6,100 Oil 0 0 0 6.05E67,101 1.21E7 1.21E7 Nitrogen 0 744 0 0 0 0 0 Facility Waste 0 0 0 0 0 00 Total 31,084 744 4,034 6.05E6 32,135 1.21E7 1.21E7 Temp., ° C. 0 25150 70 70 70 69 steam Flow, liter/hr 1,971 68,160 460,000 8.65E5 2,0381.73E6 1.73E6

TABLE 4.9 Section 2 Stream Number, Metric Tons per Year (Te/yr)Component 8 9 10 11 12 13 UO₂(NO₃)₂ 0 0 0 0 0 0 U (complexed) 0 0 0 0 00 Urea 0 0 0 0 0 0 HMTA 0 0 0 0 0 0 Water 12,100 0 6,050 6,050 4,034 0Oil 1.21E7 2 6.05E6 6.05E6 0 0 Nitrogen 0 744 0 0 0 0 Facility Waste 0 00 0 0 2,880⁽⁵⁾ Total 1.21E7 746 6.05E6 6.05E6 4,034 2,880⁽⁵⁾ Temp., ° C.69 25 69 69 150 25 Flow, liter/hr 1.73E6 68,160 8,65E5 8,65E5 4612,880⁽⁵⁾

TABLE 4.9a Section 1 Component/ Stream Number, Metric Tons per Year(Te/yr) Item 1 2 3 4 5 6 7 UO₂(NO₃)₂ 6,959 0 0 0 6,959 0 0 U (complexed)1,401 0 0 0 1,401 0 0 Urea 1,767 0 0 0 1,767 0 0 HMTA 4,963 0 0 0 4,9630 0 Carbon 855 0 0 0 855 0 0 Water 15,994 0 4,034 <1,000 9,944 6,050−6,100 Oil 0 0 0 06.05E6 7,101 1.21E7 1.21E7 Nitrogen 0 744 0 0 0 0 0Facility Waste 0 0 0 0 0 0 0 Total 31,939 744 4,034 6.05E6 32,990 1.21E71.21E7 Temp., ° C. 0 25 150 70 70 70 69 (steam) Flow, liter/hr 1.97168,160 460,000 8.65E5 2,038 1.73E6 1.73E6

TABLE 4.9a Section 2 Component/ Stream Number, Metric Tons per Year(Te/yr) Item 8 9 10 11 12 13 UO₂(NO₃)₂ 0 0 0 0 0 0 U (complexed) 0 0 0 00 0 Urea 0 0 0 0 0 0 HMTA 0 0 0 0 0 0 Carbon 0 0 0 0 0 0 Water 12,100 06,050 6,050 4,034 0 Oil 1.21E7 2 6.05E6 6.05E6 0 0 Nitrogen 0 744 0 0 00 Facility Waste 0 0 0 0 0 2,880⁽³⁾ Total 1.21E7 746 6.05E6 6.05B6 4,0342,880⁽³⁾ Temp., ° C. 69 25 69 69 150 25 Flow, l.hr 1.73E6 68,160 8.65E58.65E5 461 2,880⁽³⁾

The gel formation column for the 300 micron spheres is sized in the samemanner as the 1,200 micron spheres, only the nozzle flow rates aredifferent the values set forth above for the 1,200 micron case.

For a 300 micron final diameter, the green, gel sphere diameter would beapproximately 900 microns (0.09 cm). This results in a terminal velocityof approximately 18.3 cm/sec. Assuming the same droplet velocity as inthe 1,200 micron column (6.6 cm/sec), then the upward oil velocity wouldhave to be 14.7 cm/sec. This translates into an oil flow rate of 31,000liters per hour (136 gpm) and 86,300 liters per hour (380 gpm) for the31 and 51 cm columns, respectively. The number of nozzles and theaqueous feedrates would be the same as for the 1,200 micron case. Forsimplicity and to bound the case, the columns for the 300 microndiameter spheres are the same height as the columns for the 1,200 micronmicrospheres. A 51 cm column diameter is used as the basis because itprovides similar oil flow rate characteristics as for the 1,200 microncolumns, and, thus, would require the same sized equipment. This columnsize translates into 19 nozzles.

I. Oil Purification

FIG. 4.10 depicts the flow sheet for the Oil Purification Station. Thematerial and energy balances for the exemplary process for production ofuranium dioxide microspheres and uranium carbide microspheres are shownin Table 4.10 and Table 4.10a, respectively.

TABLE 4.10 Section 1 Stream Number, Metric Tons per Year (Te/yr)Component 1 2 3 4 5 6 7 Oil 3.35E7 6.05E6 21,667 3.96E7 3.96E7 3.96E73.96E7 Water 33,450 6,050 39,522 39,522 39,522 39,522 <6,000 FacilityWaste 0 0 0 0 0 0 0 Total 3.35E7 6.05E6 21,689 3.96E7 3.96E7 3.96E73.96E7 Temp., ° C. 69 69 25 69 2 0 0 Flow, liter/hr 4.8E6 8.65E5 3,0955.73E6 5.73E6 5.73E6 5.73E6

TABLE 4.10 Section 2 Stream Number, Metric Tons per Year (Te/yr)Component 8 9 10 11 12 13 14 Oil 4 1,156 3.96E7 3.96E7 3.35E7 6.05E6 0Water 35,550 0 <6,000 <6,000 <5,000 <1,000 0 Facility Waste 0 0 0 0 0 02,880⁽⁶⁾ Total 35,554 1,156 3.96E7 3.96E7 3.35E7 6.05E6 2,880⁽⁶⁾ Temp.,° C. 5 25 70 70 70 70 25 Flow, liter/hr 4,059 165 5.73E6 5.73E6 4.8E68.65E5 2,880⁽⁶⁾

TABLE 4.10a Section 1 Stream Number, Metric Tons per Year (Te/yr)Component 1 2 3 4 5 6 7 Oil 3.35E 6.05E6 21,667 3.96E 3.96E7 3.96E73.96E7 7 7 Water 33,450 6,050 39,522 39,522 39,522 39,522 <6,000Facility Waste 0 0 0 0 0 0 0 Total 3.35E 6.05E6 21,689 3.96E 3.96E73.96E7 3.96E7 7 7 Temp., ° C. 69 69 25 69 2 0 0 Flow, liter/hr 4.8E68.65E5 3,095 5.73E 5.73E6 5.73E6 5.73E6 6

TABLE 4.10a Section 2 Component/ Stream Number, Metric Tons per Year(Te/yr) Item 8 9 10 11 12 13 14 Oil 4 1,156 3.96E7 3.96E7 3.35E7 6.05E60 Water 35,550 0 <6,000 <6,000 <5,000 <1,000 0 Facility Waste 0 0 0 0 00 2,880⁽⁶⁾ Total 35,554 1,156 3.96E7 3.96E7 3.35E7 6.05E6 2,880⁽⁶⁾Temp., ° C. 5 25 70 70 70 70 25 Flow, liter/hr 4,059 165 5.73E6 5.73E64.8E6 8.65E5 2,880⁽⁶⁾

As discussed above, the oil used in the Gel Formation Columns willabsorb water from the gel spheres, up to the solubility limit of waterin the oil. While exact limits are not defined, gel formation precedesmore advantageously with oils that are not water saturated. Therefore,it is desirable to dry the oil.

Drying can be accomplished by several methods. Chilling of the oilfollowed by phase separation is the most straightforward method, butrequires equipment and piping. Alteratively, molecular sieves and otheradsorbents could be used to remove water from the oil, but would requirea regeneration system and handling of the tramp oil. Membrane systemscould also be used, but are probably not cost effective. Therefore, inthe preferred embodiment, chilling followed by phase separation is usedto dry the oil.

The wet oil stream 1 from the 1,200 Micron Gel Formation Station, thewet oil stream 2 from the 300 Micron Gel Formation Station, and the wetoil stream 3 from the Ammonium Hydroxide Setting/Washing Station,described below, are fed to the Oil Purification Station. Wet oilstreams 1-3 will likely contain around 1,000 ppm dissolved water, atypical value for oils around 70° C. The wet oil streams 1-3 arecombined into a single stream 4 which is introduced into a HeatExchanger and cooled. The oil 5 is then passed through a Chilling Systemand cooled to around 5° C. At this low temperature, the water solubilityis only about 100 ppm. As a result, phase separation occurs. Theresulting oil/water mixture 6 is sent to an Oil/Water Separator whichcoalesces and recovers the water. The dried oil 7 is reheated using theheat exchanger. Makeup oil 9 is added to the heated, dried oil. Theresulting mixture 11 is divided into a first portion 12 that is returnedto the 1,200 Micron Gel Formation Station and a second portion 13 thatis returned to the 300 Micron Gel Formation Station. The water 8, whichis saturated with oil, is pumped to Waste Treatment.

J. Gel Sphere Aging By Setting/Washing With Ammonium Hydroxide

FIG. 4.11 depicts the flow sheet for the 1200 MicroshpereSetting/Washing Station. The material and energy balances for theexemplary process for production of uranium dioxide microspheres anduranium carbide microspheres are shown in Table 4.11 and Table 4.11a,respectively.

TABLE 4.11 Section 1 Component/ Stream Number, Metric Tons per year(Te/yr) Item 1 2 3 4 5 6 7 UO₂(NO₃)₂ 16,238 0 0 0 0 0 0 U 3,270 0 0 0 00 0 (complexed) (NH₄)₂U₂O₇ 0 0 0 0 0 0 0 Urea 4,123 0 0 0 0 3,917 3,917HTMA 11,579 0 0 0 0 11,000 11,000 NH₄OH 0 0 14,717 14,717 0 12,15212,152 Oil 15,706 0 0 0 0 16 16 Water 3,869 0 132,451 0 129,124 9129,124 HNO₃ 0 0 0 0 0 12,454 2,454 Nitrogen 0 4,462 0 0 4,462 0 0Facility 0 0 0 0 0 0 0 Waste Total 54,785 4,462 147,168 147,168 4,462158,663 158,663 Temp., ° C. 70 25 25 25 25 25 25 Flow, liter/hr 3,474409,000 16,800 16,800 404,000 18,112 18,112

TABLE 4.11 Section 2 Component/ Stream Number, Metric Tons per Year(Te/yr) Item 8 9 10 11 UO₂(NO₃)₂ 0 0 0 0 U 0 0 0 0 (complexed)(NH₄)₂U₂O₇ 0 17,160 0 0 Urea 0 206 0 0 HTMA 0 579 0 0 NH₄OH 1 640 0 0Oil 0 769 14,921 0 Water 9 6,796 14 0 HNO₃ 0 129 0 0 Nitrogen 0 0 0 0Facility 0 0 0 2,880⁽³⁾ Waste Total 10 26,279 14,935 2,880⁽³⁾ Temp., °C. 25 25 25 25 Flow, liter/hr 1.1 1,667 2,131 2,880⁽³⁾

TABLE 4.11a Section 1 Component/ Stream Number, Metric Tons per Year(Te/yr) Item 1 2 3 4 5 6 7 UO₂(NO₃)₂ 16,238 0 0 0 0 0 0 U 3,270 0 0 0 00 0 (complexed) (NH₄)₂U₂O₇ 0 0 0 0 0 0 0 Urea 4,123 0 0 0 0 3,917 3,917HTMA 11,579 0 0 0 0 11,000 11,000 NH₄OH 0 0 14,717 14,717 0 16 16 Oil15,706 0 0 0 0 16 16 Carbon 1,995 0 0 0 0 0 0 Water 3,869 0 132,451 0129,124 9 129,124 HNO₃ 0 0 0 0 0 12,454 2,454 Nitrogen 0 4,462 0 0 4,4620 0 Facility 0 0 0 0 0 0 0 Waste Total 56,780 4,462 147,168 147,1684,462 158,66 158,663 3 Temp., ° C. 70 25 25 25 25 25 25 Flow, liter/hr3,474 409,000

TABLE 4.11a Section 2 Component/ Stream Number, Metric Tons per Year(Te/yr) Item 8 9 10 11 UO₂(NO₃)₂ 0 0 0 0 U (complexed) 0 0 0 0(NH₄)₂U₂O₇ 0 17,160 0 0 Urea 0 206 0 0 HTMA 1 579 0 0 NH₄OH 0 640 0 0Oil 0 769 14,921 0 Carbon 0 1,995 0 0 Water 9 6,796 14 0 HNO₃ 0 129 0 0Nitrogen 0 0 0 0 Facility Waste 0 0 0 2,880⁽³⁾ Total 10 28,274 14,9352,880⁽³⁾ Temp., ° C. 25 25 25 25 Flow, liter/hr 1.1 1,667 2,131 2,880⁽³⁾

Additional ammonia necessary to complete the ADU-like precipitationreaction is supplied by washing the spheres in a setting solution of 10%ammonium hydroxide solution. The stoichiometry for this reaction is:

 2UO₂(NO₃)₂+2NH₄OH+H₂O→(NH₄)₂U₂O₇+4HNO₃

During this process, the spheres 1 are periodically discharged from the1,200 Micron Gel Formation Column into a Setting Washing Column havingthe same diameter as the formation column (31 cm). A setting solution 4of 10% ammonium hydroxide is formed by combining the ammonium hydroxidestream 3 from the Ammonium Hydroxide Purification Station, describedbelow, and a portion 8 of the ammonium hydroxide 6 from the AmmoniumHydroxide Overflow Fank. The setting solution 4 is circulated throughthe bed of spheres in the Setting/Washing Column, causing them toharden. Typical washing times are about one hour. Typical flow rates areon the order of five to ten times the initial broth solution feed rateto the column. Ambient temperatures are believed to supply sufficientheat for the reaction.

During washing, impurities such as urea and HMTA are removed from thespheres by leaching (i.e., preferential absorption into the aqueousphase), typically with efficiencies of 90-95%, because of favorablethermodynamics.

About 95% of the tramp oil carried over from the Gel Formation Stationis skimmed from the Ammonium Hydroxide Overflow Tank. This oil 10 isthen recycled to the Oil Purification Station. A portion of the oil isbelieved to dissolve to ammonium hydroxide solution, approximately 100ppm, while the remainder clings to the spheres. The ammonium hydroxidealso causes reformation of any HMTA that initially reacted with the ADU.

After washing, a majority 7 of the ammonium hydroxide 6 collected in theAmmonium Hydroxide Overflow Tank is sent to the Ammonium HydroxidePurification System to remove impurities (i.e., the urea and HMTA) andto allow its recycle. The purified ammonium hydroxide solution 3 is thenreturned to the 1,200 Micron Sphere Setting/Washing Station and combinedwith a minority 8 of the ammonium hydroxide solution 8 collected in thefrom the Ammonium Hydroxide Overflow Tank.

As set forth in the material and energy balances in Table 4.11, it isbelieved that the remaining 95% of the precipitation reaction occurs inthis step. Furthermore, oil carryover is believed to be equivalent tothe void space in the packed bed volume of the spheres. Therecirculation rate is calculated based upon five times the brothsolution feed to a 31 cm diameter column; this corresponds toapproximately 2.3 liters per minute (0.62 gpm) per column.

After setting and washing of the gel spheres is complete (i.e., theuranium is completely converted to ADU and no more ammonium hydroxidereacts), the aged spheres 9 are transferred to the 1,200 Micron DryingStation for further processing.

Facility waste 11 from the 1,200 Micron Sphere Setting/Washing Stationis sent to Waste Treatment.

FIG. 4.12 depicts the flow sheet for the 300 Microshpere Setting/WashingStation. The material and energy balances for the exemplary process forproduction of uranium dioxide microspheres and uranium carbidemicrospheres are shown in Table 4.12 and Table 4.12a, respectively.

TABLE 4.12 Section 1 Component/ Stream Number, Metric Tons per Year(Te/yr) Item 1 2 3 4 5 6 7 UO₂(NO₃)₂ 6,959 0 0 0 0 0 0 U (complexed)1,401 0 0 0 0 0 0 (NH₄)₂U₂O₇ 0 0 0 0 0 0 0 Urea 1,767 0 0 0 0 1,6791,679 HTMA 4.963 0 0 0 0 4,715 4,715 NH₄OH 0 0 6,658 6,658 0 5,541 5,514Oil 7,101 0 0 0 0 8 8 Water 9,944 0 59,918 59,918 0 66,164 66,164 HNO₃ 00 0 0 0 1,410 1,410 Nitrogen 0 744 0 0 744 0 0 Facility Waste 0 0 0 0 00 0 Total 32,135 744 66,576 66,576 744 79,4907 79,490 Temp., ° C. 70 2525 25 25 25 25 Flow, liter/hr 2,038 68,160 7,600 68,160 9,074 9,074 1.1

TABLE 4.12 Section 2 Component/ Stream Number, Metric Tons per Year(Te/yr) Item 8 9 10 11 UO₂(NO₃)₂ 0 0 0 0 U (complexed) 0 0 0 0(NH₄)₂U₂O₇ 0 7,488 0 0 Urea 0 88 0 0 HTMA 0 248 0 0 NH₄OH 1 292 0 0 Oil0 329 6,746 0 Water 9 3,482 7 0 HNO₃ 0 74 0 0 Nitrogen 0 0 0 0 FacilityWaste 0 0 0 2,880 Total 10 12,001 6,753 2,880 Temp., ° C. 25 25 25 25Flow, liter/hr 1.1 761 964 2,880

TABLE 4.12a Section 1 Component/ Stream Number, Metric Tons per Year(Te/yr) Item 1 2 3 4 5 6 7 UO₂(NO₃)₂ 6,959 0 0 0 0 0 0 U (complexed)1,401 0 0 0 0 0 0 (NH₄)₂U₂O₇ 0 0 0 0 0 0 0 Urea 1,767 0 0 0 0 1,6791,679 HTMA 4,963 0 0 0 0 4,715 4,715 NH₄OH 0 0 6,658 6,658 0 5,541 5,541Oil 7,101 0 0 0 0 8 8 Carbon 855 0 0 0 0 0 0 Water 9,944 0 59,918 59,9180 66,164 66,164 HNO₃ 0 0 0 0 0 1,410 1,410 Nitrogen 0 744 0 0 744 0 0Facility Waste 0 0 0 0 0 0 0 Total 32,990 744 66,576 66,576 744 79,49079,490 Temp., ° C. 70 25 25 25 25 25 25 Flow, liter/hr 2,038 68,1607,600 7,600 68,160 9,074 9,074

TABLE 4.12a Section 2 Component/ Stream Number, Metric Tons per Year(Te/yr) Item 8 9 10 11 UO₂(NO₃)₂ 0 0 0 0 U (complexed) 0 0 0 0(NH₄)₂U₂O₇ 0 7,488 0 0 Urea 0 88 0 0 HTMA 0 248 0 0 NH₄OH 1 292 0 0 Oil0 329 6,746 0 Carbon 0 855 0 0 Water 9 3,482 7 0 HNO₃ 0 74 0 0 Nitrogen0 0 0 0 Facility Waste 0 0 0 2,880⁽³⁾ Total 10 12,856 6,753 2,880⁽³⁾Temp., ° C. 25 25 25 25 Flow, liter/hr 1.1 761 964 2,880⁽³⁾

The 300 Micron Setting/Washing Station is identical to the 1,200 MicronSetting/Washing Station. As with the 1,200 micron spheres, the remainderof the ammonia required to complete the ADU-like precipitation reactionis supplied by washing the spheres in a setting solution of 10% ammoniumhydroxide. Furthermore, as with the 1,200 micron particles, typicalwashing times are on the order of one hour and typical flow rates arearound five times the flow of the initial broth solution fed to thecolumn. These conditions correspond to a large excess of ammoniumhydroxide, since only around 10% or so is actually consumed by thesetting reactions. Hence, the reactions are rapid.

K. Gel Sphere Drying And Liberation Of Ammonia and Water

FIG. 4.13 depicts the flow sheet for the 1200 Microshpere DryingStation. The material and energy balances for the exemplary process forproduction of uranium dioxide microspheres and uranium carbidemicrospheres are shown in Table 4.13 and Table 4.13a, respectively.

TABLE 4.13 Section 1 Component/ Stream Number, Metric Tons per Year(Te/yr) Item 1 2 3 4 5 6 7 (NH₄)₂U₂O₇ 17,160 0 0 0 0 17,160 0 Urea 206 00 0 0 206 0 HTMA 579 0 0 0 0 579 0 NH₄OH 640 608 0 608 0 32 0 Oil 769 00 0 0 769 0 Nitrogen 0 711,513 2,231 711,513 7,115 0 711,513 Water 6,7966,456 0 6,456 0 340 0 HNO₃ 129 123 0 123 0 6 0 Facility Waste 0 0 0 0 00 Total 26,279 718,700 2,231 718,700 7,115 19,092 711,513 Temp., ° C. 2530 25 30 25 100 120 Flow, liter/hr 1,667 6.56E7 409,000 6.56E7 6.56E71,211 6.55E7

TABLE 4.13 Section 2 Component/ Stream Number, Metric Tons per Year(Te/yr) Item 8 9 10 (NH₄)₂U₂O₇ 0 0 0 Urea 0 0 0 HTMA 0 0 0 NH₄OH 0 608 0Oil 0 0 0 Nitrogen 7,115 0 0 Water 65 6,341 0 HNO₃ 0 123 0 FacilityWaste 0 0 2,880(4) Total 7,180 7,122 2,880(4) Temp., ° C. 25 25 25 Flow,liter/hr 6.56E5 813 2,880(4)

TABLE 4.13a Section 1 Component/ Stream Number, Metric Tons per Year(Te/yr) Item 1 2 3 4 5 6 7 (NH₄)₂U₂O₇ 17,160 0 0 0 0 17,160 0 Urea 206 00 0 0 206 0 HTMA 579 0 0 0 0 579 0 NH₄OH 640 608 0 608 0 32 0 Oil 769 00 0 0 769 0 Nitrogen 0 711,513 2,231 711,513 7,115 0 711,513 Carbon1,995 0 0 0 0 1,995 0 Water 6,796 6,456 0 6,456 0 340 0 HNO₃ 129 123 0123 0 6 0 FacilityWaste 0 0 0 0 0 0 0 Total 28,274 718,700 2,231 718,7007,115 21,087 711,513 Temp., ° C. 25 30 25 30 25 100 120 Flow, liter/hr1,667 6.56E7 409,000 6.56E7 6.56E7 1,211 6.55E7

TABLE 4.13a Section 2 Component/ Stream Number, Metric Tons per Year(Te/yr) Item 8 9 10 (NH₄)₂U₂O₇ 0 0 0 Urea 0 0 0 HTMA 0 0 0 NH₄OH 0 608 0Oil 0 0 0 Nitrogen 7,115 0 0 Carbon 0 0 0 Water 65 6,341 0 HNO₃ 0 123 0Facility Waste 0 0 2,880(4) Total 7,180 7,122 2,880(4) Temp., ° C. 25 2525 Flow, liter/hr 6.56E5 813 2,880(4)

In the preferred embodiment, the spheres are subjected to slow, lowtemperature drying in order to remove the volatile species, such aswater, nitric acid, and ammonia, while avoiding cracking. Non-volatilespecies, such as urea and HMTA, are not removed in this process.Preferably, the aged spheres 1 from the 1,200 Micron SphereSetting/Washing Station are introduced into a commercial Tray Dryer. TheTray Dryer is equipped with a moving screen which provides a contactingarrangement for drying with warm nitrogen gas 7. Warm nitrogen gas 7 ispurified in a commercial Nitrogen Condenser/Dryer and recycled.Adiabatic drying is assumed.

The Nitrogen Condenser/Dryer uses refrigeration to condense volatilespecies 9 which are transferred to the Ammonium Hydroxide PurificationStation. A recuperative heat exchanger (not shown) minimizes energyconsumption. An adsorbent bed of molecular sieves (not shown) is alsoused to further dry the nitrogen. The molecular sieves are periodicallyregenerated by a purge air stream (not shown) The material and energybalances for the exemplary process in Table 4.13 utilize 99%condensation of the water in the Tray Dryer and 100% condensation of thenitric acid and ammonium hydroxide in the water. The remaining 1% of thewater is removed by the adsorbent bed. Although leak-free operation hasbeen attained commercially, the analyses in Table 4.13 utilize anitrogen makeup stream 4 represents the reconstituted nitrogen stream.Stream 5 is the supply air used to regenerate and purge the dryer, andbecomes stream 8.

The dried spheres 6 are removed from the Tray Dryer and transferred tothe 1,200 Micron Sphere Conversion/Sintering Station.

Facility waste 10 from the 1,200 Micron Sphere Drying Station is sent toWaste Treatment.

FIG. 4.14 depicts the flow sheet for the 300 Microshpere Drying Station.The material and energy balances for the exemplary process forproduction of uranium dioxide microspheres and uranium carbidemicrospheres are shown in Table 4.14 and Table 4.14a, respectively.

TABLE 4.14 Section 1 Component/ Stream Number, Metric Tons per Year(Te/yr) Item 1 2 3 4 5 6 7 (NH₄)₂U₂O₇ 7,488 0 0 0 0 7,488 0 Urea 88 0 00 0 88 0 HTMA 248 0 0 0 0 248 0 NH₄OH 292 277 0 277 0 15 0 Oil 329 0 0 00 329 0 Nitrogen 0 361,845 372 361,845 3,618 0 361,845 Water 3,482 3,3080 3,308 0 174 0 HNO₃ 74 70 0 70 0 4 0 Facility Waste 0 0 0 0 0 0 0 Total12,001 365,500 372 365,500 3,618 8,346 361,845 Temp., ° C. 25 30 30 3025 100 120 Flow, liter/hr 761 3.34E7 68,200 3.34E7 330,000 529 3.34E7

TABLE 4.14 Section 2 Component/ Stream Number, Metric Tons per Year(Te/yr) Item 8 9 10 (NH₄)₂U₂O₇ 0 0 0 Urea 0 0 0 HTMA 0 0 0 NH₄OH 0 277 0Oil 0 0 0 Nitrogen 3,618 0 0 Water 33 3,275 0 HNO₃ 0 70 0 Facility Waste0 0 2,880(4) Total 3,651 3,622 2,880(4) Temp., ° C. 25 25 25 Flow,liter/hr 330,000 400 2,880(4)

TABLE 4.14a Section 1 Component/ Stream Number, Metric Tons per Year(Te/yr) Item 1 2 3 4 5 6 7 (NH₄)₂U₂O₇ 7,488 0 0 0 0 7,88 0 Urea 88 0 0 00 88 0 HTMA 248 0 0 0 0 248 0 NH₄OH 292 277 0 277 0 15 0 Oil 329 0 0 0 0329 0 Nitrogen 0 361,845 372 361,845 3,618 0 361,845 Carbon 855 0 0 0 0855 0 Water 3,482 3,308 0 3,308 0 174 0 HNO₃ 74 70 0 70 0 4 0 FacilityWaste 0 0 0 0 0 0 0 Total 12,85 365,500 372 365,500 3,618 9,201 361,8456 Temp., ° C. 25 30 30 30 25 100 120 Flow, liter/hr 761 3.34E7 68,2003.34E7 330,000 529 3.34E7

TABLE 4.14a Section 2 Component/ Stream Number, Metric Tons per Year(Te/yr) Item 8 9 10 (NH₄)₂U₂O₇ 0 0 0 Urea 0 0 0 HTMA 0 0 0 NH₄OH 0 277 0Oil 0 0 0 Nitrogen 3,618 0 0 Carbon 0 0 0 Water 33 3.275 0 HNO₃ 0 70 0Facility Waste 0 0 2,880(4) Total 3,651 3,622 2,880(4) Temp., ° C. 25 2525 Flow, liter/hr 330,0 400 2,880(4) 00

This station performs the drying for the 300 micron sized product. Apartfrom the flow rates, the operations are identical to those for the 1,200micron spheres discussed in Section 4.4.14. FIG. 4.14 shows the flowsheet, and Table 4.14 summarized the mass and energy balances.

L. Gel Sphere Conversion and Sintering By Heating

FIG. 4.15 depicts the flow sheet for the 1200 Microsphere Conversion andSintering/Washing Station. The material and energy balances for theexemplary process for production of uranium dioxide microspheres anduranium carbide microspheres are shown in Table 4.15 and Table 4.15a,respectively.

TABLE 4.15 Component/ Stream Number, Metric Tons per Year (Te/yr) Item 12 3 4 5 (NH₄)₂U₂O₇ 17,160 0 0 0 0 UO₂ 0 0 14,850 0 0 NH₃ 0 0 0 1,350 0CO₂ 0 0 0 3,660 0 N₂ 0 0 0 Trace 0 Argon 0 10,725 0 10,725 0 Hydrogen 0110 0 429 0 NO₂ 0 0 0 4.6 0 Urea 206 0 0 0 0 HTMA 579 0 0 0 0 NH₄OH 32 00 0 0 Oil 769 0 0 0 0 Nitrogen 0 0 0 0 0 Water 340 1,100 0 0 0 FacilityWaste 0 0 0 0 0 Total 19,092 11,935 14,850 16,169 0 Temp., ° C. 25 150100 300 25 Flow, liter/hr 1,667 1,16E6 188 2.2E6 0

TABLE 4.15 Component/ Stream Number, Metric Tons per Year (Te/yr) Item 12 3 4 5 (NH₄)₂U₂O₇ 17,160 0 0 0 0 UC₂ 0 0 14,588 0 0 NH₃ 0 0 0 1,350 0CO₂ 0 0 0 5,636 0 N₂ 0 0 0 Trace 0 Argon 0 10,725 0 10,725 0 Hydrogen 0110 0 429 0 NO₂ 0 0 0 4.6 0 Urea 206 0 0 0 0 HTMA 579 0 0 0 0 NH₄OH 32 00 0 0 Oil 769 0 0 0 0 Nitrogen 0 0 0 0 0 Carbon 1,995 0 0 0 0 Water 3401,100 0 0 0 Facility Waste 0 0 0 0 0 Total 21,087 11,935 14,588 18,145 0Temp., ° C. 25 150 100 300 25 Flow, liter/hr 1,667 1.16E6 146 2.5E6 0

The dried spheres 1 obtained from the 1,200 Micron Drying Station areintroduced into a Vertical Tube Furnace (VTF) where the spheres areconverted to uranium dioxide and sintered (i.e., densified). The driedspheres react in the VTF according to the following equation:

(NH₄)₂U₂O₇+2H₂→2UO₂+3H₂O+2NH₃

The VTF consists of a vertical tube containing a slowly moving packedbed of ADU spheres within an electrically heated furnace. Dry spheres 1are added at the top of the VTF, while a discharge valve dischargesdense product 3 from the bottom.

Typical residence times in the VTF approach twelve hours at maximumtemperatures of 1,100 to 1,300° C. In normal conversion and sintering,hydrogen is added to an argon diluent, typically at concentrations of3.5-4 ppmv, and the gas mixture 2 converted to uranium dioxide andsintered (i.e., densified) is circulated upwards through the VTF.Approximately 50% of the hydrogen is consumed in this process, and thesurviving off gas 4 is sent to the Gas Purification Station for recycle.The concentrations of hydrogen in the argon/hydrogen mixture are belowthe lower flammability limit for hydrogen in air, and, thus, preventhydrogen fire hazards within the VTF system itself.

Meanwhile, urea, HMTA, ammonium hydroxide, nitric acid, and the oilthermally crack and reform in the VTF as follows:

CO(NH₂)₂+H₂O→CO₂+2NH₃(urea)

(CH₂)₆N₄+12H₂O→4NH₃+12H₂+6CO₂(HMTA)

NH₄OH→NH₃+H₂O

2HNO₃+H₂→2NO₂+2H₂O

(CH₂)₁₀+20H₂O→10CO₂+30H₂

The enthalpy balance set forth in Table 4.15 for the exemplary processutilizes a value of 30 KW per VTF, based upon average values reported inthe literature. This enthalpy is approximately twice that required forvolatilization of the chemical species. For uranium dioxide flow ratesof 15,000 tonnes/year, approximately 20 VTF's are required. Unlikenormal conversion and sintering operations, there is a net hydrogenproduction generated from the cracking and reforming of the traceimpurities. Also, since the reactions result in a net consumption ofwater, steam (not shown) is added to the argon diluent in the VTF feed.Alternatively, steam could be added internally within the VTF. Thesehydrogen and carbon production effects accrue from the carryover ofresidual oil upon the spheres. The argon flowrates are based uponmaintaining a 4% hydrogen in argon concentration at the exit of the VTF.

The preferred embodiment uses nitrogen in place of argon because of itslower cost and the residual nitrogen in the dense uranium (as uraniumnitride) provides additional neutron shielding.

Facility waste 5 from the 300 Micron Sphere Conversion and SinteringStation is sent to Waste Treatment.

FIG. 4.16 depicts the flow sheet for the 300 Microshpere Conversion andSintering Station. The material and energy balances for the exemplaryprocess for production of uranium dioxide microspheres and uraniumcarbide microspheres are shown in Table 4.16 and Table 4.16a,respectively.

TABLE 4.16 Component/ Stream Number, Metric Tons per Year (Te/yr) Item 12 3 4 5 (NH₄)₂U₂O₇ 7,488 0 0 0 0 UO₂ 0 0 6,480 0 0 NH₃ 0 0 0 585 0 CO₂ 00 0 1,566 0 N₂ 0 0 0 Trace 0 Argon 0 4,575 0 4,575 0 Hydrogen 0 48 0 1830 NO₂ 0 0 0 5.8 0 Urea 88 0 0 0 0 HTMA 248 0 0 0 0 NH₄OH 15 0 0 0 0 Oil329 0 0 0 0 HNO₃ 4 0 0 0 0 Water 174 600 0 0 0 Facility Waste 0 0 0 0 0Total 8,346 5,223 6,480 6,915 0 Temp, 20 C. 100 150 100 300 25 Flow,liter/hr 529 5.08E5 82 9.41E5 0

TABLE 4.16a Component/ Stream Number, Metric Tons per Year (Te/yr) Item1 2 3 4 5 (NH₄)₂U₂O₇ 7,488 0 0 0 0 UO₂ 0 0 6,252 0 0 NH₃ 0 0 0 585 0 CO₂0 0 0 2,598 0 N₂ 0 0 0 Trace 0 Argon 0 4,575 0 4,575 0 Hydrogen 0 48 0183 0 NO₂ 0 0 0 5.8 0 Urea 88 0 0 0 0 HTMA 248 0 0 0 0 NH₄OH 15 0 0 0 0Oil 329 0 0 0 0 HNO₃ 4 0 0 0 0 Carbon 855 0 0 0 0 Water 174 600 0 0 0Facility Waste 0 0 0 0 0 Total 9,201 5,223 6,252 7,947 0 Temp., ° C. 100150 100 300 25 Flow, liter/hr. 529 5.08E5 63 1.08E6 0

Conversion and sintering of the 300 micron particles occur in a manneranalogous to the 1,200 micron spheres. Approximately 8 VTF's arerequired for the 300 micron spheres.

M. Gel Sphere Collection

As discussed previously, the final dense spheres are approximately 30%the size of the initial droplets (gel spheres). It is desirable to haveat least two size ranges for better packing, higher bulk densities, and,consequently, better shielding. These size ranges are 1-2 mm for thecoarse fraction and 0.030-1 mm for the fine fraction. The coarse spherescontain the majority of the uranium on a mass basis.

N. Calcium Nitrate Reconstitution for the Calcium Nitrate ReconstitutionStation

FIG. 4.17 depicts the flow sheet for the Calcium Nitrate ReconstitutionStation. The material and energy balances for the exemplary process forproduction of uranium dioxide microspheres and uranium carbidemicrospheres are shown in Table 4.17 and Table 4.1 7a, respectively.

TABLE 4.17 Component/ Stream Number, Metric Tons per Year (Te/yr) Item 12 3 4 5 6 HNO₃ 6,804 0 3,110 0 0 0 Lime (CaO) 0 4,396 0 0 0 0 Ca(NO₃₎ ₂0 0 0 12,874 0 12,874 Water 2,268 0 141,214 144,815 93,399 51,496 Total9,072 4,396 144,324 157,769 93,399 64,370 Temp., ° C. 25 0 60 60 60 60Flow, l/hr 1,036 335 16,475 18,010 10,662 6,100

TABLE 4.17a Component/ Stream Number, Metric Tons per Year (Te/yr) Item1 2 3 4 5 6 HNO₃ 6,801 0 3,110 0 0 0 Lime (CaO) 0 4,396 0 0 0 0 Ca(NO₃)₂0 0 0 12,874 0 12,874 Water 2,268 0 141,214 144,815 93,399 51,496 Total9,072 4,396 144,324 157,769 93,399 64,370 Temp., ° C. 25 0 60 60 60 60Flow, l/hr 1,036 335 16,475 18,000 10,662 6,100

The overhead product 3 consisting of a dilute nitric acid stream fromthe Uranyl Nitrate Solution Adjustment Station is added to a Mixing Tankalong with dry lime 2 to form calcium nitrate. Additional nitric acid 1is added to the output 3 of the Mixing Tank, and the mixture 4 is sentto a Vapor Recompression (VR) Evaporator where Excess Water 5 is removedand sent to the Deionized Water System. The Reconstituted Product 6 fromthe VR Evaporator containing the calcium nitrate is transferred to theUranyl Nitrate Formation Station for reuse.

O. Ammonium Hydroxide Solution Purification

FIG. 4.18 depicts the flow sheet for the Ammonium Hydroxide SolutionPurification Station. The material and energy balances for the exemplaryprocess for production of uranium dioxide microspheres and uraniumcarbide microspheres are shown in Table 4.18 and Table 4.18a,respectively.

TABLE 4.18 Section 1 Component/ Stream Number, Metric Tons per Year(Te/yr) Item 1 2 3 4 5 6 7 Urea 3,917 1,679 0 0 0 5,596 5,596 HMTA11,000 4,715 0 0 0 15,715 15,715 NH₄OH 12,152 5,541 0 0 0 17,693 18,578Oil 16 8 Trace Trace 0 24 24 HNO₃ 2,454 1,410 0 0 0 3,864 0 Water129,124 66,164 0 0 2,600 195,288 208,713 NaOH (50%) 0 0 0 0 2,600 0 24NaNO₃ 0 0 0 0 0 0 5,474 Facility Waste 0 0 0 0 0 0 0 Total 158,66379,490 Trace Trace 5,200 238,180 254,124 Temp., ° C. 25 25 25 25 25 2525 Flow, liter/hr 18,112 9,074 0 0 495 27,186 24,175

TABLE 4.18 Section 2 Stream Number, Metric Tons per Year (Te/yr)Component 8 9 10 11 12 13 14 Urea 0 5,596 0 5,596 0 0 0 HMTA 0 15,715 015,715 0 0 0 NH₄OH 18,578 0 0 0 0 608 277 Oil 0 24 0 0 0 0 0 HNO₃ 0 0 00 0 123 70 Water 167,202 41,511 8,505 33,006 0 6,391 3,275 NaOH (50%) 024 24 0 0 0 0 NaNO₃ 0 5,474 5,474 0 0 0 0 Facility 0 0 0 0 2,880 0 0Waste Total 185,780 68,344 14,003 54,317 2,880 7,122 3,622 Temp., ° C.40 40 40 40 25 25 25 Flow, liter/hr 17,700 6,500 1,332 5,200 2,880 813400

TABLE 4.18 Section 3 Stream Number, Metric Tons per Year (Te/yr)Component 15 16 17 Urea 0 5,596 5,596 HTMA 0 15,715 15,715 NH₄OH 88518,578 0 Oil 0 24 0 HNO₃ 193 4,057 0 Water 9,666 204,954 41,511 NaOH(50%) 0 0 24 NaNO₃ 0 0 5,474 Facility 0 0 0 Waste Total 10,744 248,92468,320 Temp., ° C. 25 25 25 Flow, liter/hr 1,213 23,680 6,500

TABLE 4.18a Stream Number, Metric Tons per Year (Te/yr) Component 1 2 34 5 6 Urea 3,917 1,679 0 0 0 5,596 HMTA 11,000 4,715 0 0 0 15,715 NH₄OH12,152 5,541 0 0 0 17,693 Oil 16 8 Trace Trace 0 24 HNO₃ 2,454 1,410 0 00 3,864 Water 129,12 66,164 0 0 2,600 195,288 4 NaOH (50%) 0 0 0 0 2,6000 NaNO₃ 0 0 0 0 0 0 Facility Waste 0 0 0 0 0 0 Total 158,66 79,490 TraceTrace 5,200 238,180 3 Temp., ° C. 25 25 25 25 25 25 Flow, liter/hr18,112 9,074 0 0 495 27,186

Ammonium hydroxide solution from the 1,200 Micron Setting/WashingStation 1 and from the 300 Micron Setting/Washing Station 2 are sent toOil/Water Separators, which remove any excess oil from the respectiveammonium hydroxide solutions. The separated oil portions 3 and 4 aresent to the 1,200 Micron and 300 Micron Gel Forming Stations,respectively. In a large plant, significant quantities of oil would beexpected to be carried over due to the size of the equipment. Next, theammonium hydroxide streams from the Oil/Water Separators are merged intoa single stream 6 and combined with the condensate 15 formed by thecombination of the condensates 13 and 14 from the 1,200 Micron DryerStation and the 300 Micron Dryer Station, respectively. A 50% SodiumHydroxide Solution 5 is added to the condensate/ammonium hydroxidemixture 16 to neutralize the nitric acid, displace ammonia, and raisethe solution pH above 10. A VR Evaporator processes the resultingmixture 7. The overhead stream 8 from the VR Evaporator is a dilute, 10%ammonium hydroxide solution, which is sent to the Ammonium HydroxideSolution Reconstitution Station, described below, for furtherprocessing. The bottoms 9 from the UR Evaporator contain sodium nitrate,urea, and HMTA, with a total dissolved solids concentration approaching40%. Traces of oil are removed from the bottoms 9 by activated carbon ina Carbon Guard Bed. A membrane system, such as a reverse osmosismembrane, separates sodium nitrate permeate 10 from the urea and HMTA inthe de-oiled solution 17. The sodium nitrate permeate 10 is sent toWaste Treatment, while the retentate 11 reports to the Urea and HMTARecovery Station, described below. An optimized design for this portionof the system would preferably evaporate less water and use solubilitylimits to the utmost advantage. Facility waste 12 from the AmmoniumHydroxide Solution Purification Station is sent to Waste Treatment.

P. Vertical Tube Furnace Gas Purification

FIG. 4.19 depicts the flow sheet for the Vertical Tube Furnace GasPurification Station. The material and energy balances for the exemplaryprocess for production of uranium dioxide microspheres and uraniumcarbide microspheres are shown in Table 4.19 and Table 4.19a,respectively.

TABLE 4.19 Section 1 Component/ Stream Number, Metric Tons per Year(Te/yr) Item 1 2 3 4 5 6 7 UO₂ 0 0 0 0 0 0 0 NH₃ 0 1,350 585 1,935 0 0 0CO₂ 0 3,660 1,566 5,226 0 0 0 N₂ 0 Trace Trace Trace 0 0 0 Argon 010,725 4,575 15,300 0 0 0 Hydrogen 0 429 183 612 0 0 0 NO₂ 0 4.6 5.810.4 0 0 0 NaOH 0 0 0 0 0 0 0 Na₂CO₃ 0 0 0 0 0 0 0 NH₄OH 0 0 0 0 0 0 0HNO₃ 0 0 0 0 0 0 0 Water 1700 0 0 0 90,707 43,603 47,104 Facility Waste0 0 0 0 0 0 0 Total 1,700 16,169 6,915 23,084 90,707 43,603 47,104Temp., ° C. 150 300 300 300 25 25 25 Flow, l/hr 194,000 2.05E6 8.8E52.93E6 10,355 4,978 5.377

TABLE 4.19 Section 1 Stream Number, Metric Tons per Year (Te/yr)Component 8 9 10 11 12 13 14 UO₂ 0 0 0 0 0 0 0 NH₃ 0 0 1,935 0 0 0 0 CO₂0 0 5,226 0 5.226 0 0 N₂ 0 0 0 0 0 0 0 Argon 0 0 15,300 0 15,300 0 0Hydrogen 0 0 612 0 612 612 0 NO₂ 0 0 10.4 0 0 0 0 NaOH 9,502 9,502 0 0 00 Trace Na₂CO₃ 0 0 0 0 0 0 12,614 NH₄OH 0 0 0 9,660 0 0 0 HNO₃ 0 0 014.5 0 0 0 Water 9,502 0 0 38,625 0 0 50,456 Facility Waste 0 0 0 0 0 00 Total 19,004 51,904 23,084 48,300 21,138 15,912 63,070 Temp., ° C. 2525 100 25 25 25 25 Flow, l/hr 1,808 5,925 1.96E6 5,500 1.79E6 1.4E67,200

TABLE 4.19 Section 3 Stream Number, Metric Tons per Year (Te/yr)Component 15 16 17 18 19 20 21 UO₂ 0 0 0 0 0 0 0 NH₃ 0 0 0 0 0 0 0 CO₂ 00 0 0 0 0 0 N₂ 0 0 0 0 0 0 0 Argon 0 15,300 15,300 10,725 4,575 153 0Hydrogen 454 158 158 110 48 0 0 NO₂ 0 0 0 0 0 0 0 NaOH 0 0 0 0 0 0 0Na₂CO₃ 0 0 0 0 0 0 0 NH₄OH 0 0 0 0 0 0 0 HNO₃ 0 0 0 0 0 0 0 Water 0 01,700 1,100 600 0 0 Facility Waste 0 0 0 0 0 0 2,880 Total 454 15,45817,158 11,935 5,223 153 2,880 Temp., ° C. 25 25 150 150 150 25 25 Flow,l/hr 5.8E5 1.35E6 1.6E6 1.1E6 4.7E5 9,780 2,880

TABLE 4.19a Section 1 Stream Number, Metric Tons per Year (Te/yr)Component 1 2 3 4 5 6 7 UO₂ 0 0 0 0 0 0 0 NH₃ 0 1,350 585 1,935 0 0 0CO₂ 0 5,636 2,598 8,234 0 0 0 N₂ 0 Trace Trace Trace 0 0 0 Argon 010,725 4,575 15,300 0 0 0 Hydrogen 0 429 183 612 0 0 0 NO₂ 0 4.6 5.810.4 0 0 0 NaOH 0 0 0 0 0 0 0 Na₂CO₃ 0 0 0 0 0 0 0 NH₄OH 0 0 0 0 0 0 0HNO₃ 0 0 0 0 0 0 Water 1700 0 0 0 90,707 43,603 47,104 Facility Waste 00 0 0 0 0 0 Total 1,700 18,145 7,947 26,902 90,707 43,603 47,104 Temp.,° C. 150 300 300 300 25 25 25 Flow, l/hr 194 2.05E6 1.08E6 3.58E6 10,3554,978 5,377

TABLE 4.19a Section 2 Stream Number, Metric Tons Per Year (Te/yr)Component 8 9 10 11 12 13 14 UO₂ 0 0 0 0 0 0 0 NH₃ 0 0 1,935 0 0 0 0 CO₂0 0 8234 0 8,234 0 0 N₂ 0 0 0 0 0 0 0 Argon 0 0 15,300 0 15,300 0 0Hydrogen 0 0 612 0 612 612 0 NO₂ 0 0 10.4 0 0 0 0 NaOH 14,97 14,971 0 00 0 Trace 1 Na₂CO₃ 0 0 0 0 0 0 19,874 NH₄OH 0 0 0 9,660 0 0 0 HNO₃ 0 0 014.5 0 0 0 Water 9,502 0 0 38,625 0 0 50,456 Facility Waste 0 0 0 0 0 00 Total 29,94 51,904 26,092 48,300 21,138 15,912 70,330 7 Temp., ° C. 2525 100 25 25 25 25 Flow, l/hr 2,848 5,925 2.4E6 5,500 2.04E6 1.4E6 8,030

TABLE 4.19a Section 3 Stream Number, Metric Tons Per Year (Te/yr)Component 15 16 17 18 19 20 21 UO₂ 0 0 0 0 0 0 0 NH₃ 0 0 0 0 0 0 0 CO₂ 00 0 0 0 0 0 N₂ Argon 0 15,300 15,300 10,725 4,575 153 0 Hydrogen 454 158158 110 48 0 0 NO₂ NaOH 0 0 0 0 0 0 0 Na₂CO₃ 0 0 0 0 0 0 0 NH₄OH 0 0 0 00 0 0 HNO₃ 0 0 0 0 0 0 0 Water 0 0 1,700 1,100 600 0 0 Facility Waste 00 0 0 0 0 2,880 Total 454 15,458 17,158 11,935 5,223 153 2,880 Temp., °C. 25 25 150 150 150 25 25 Flow, l/hr 5.8E5 1.35E6 1.6E6 1.1E6 4.7E59,780 2,880

The hot gases 2 and 3 from the VTF's in the 1,200 Micron and 300 MicronConversion and Sintering Station, respectively, are merged into a singlestream 4 and filtered. (Although it is expected that little or nomaterial will be captured due to the size, narrow diameter distribution,and density of the microspheres.) The filtered gas is then passedthrough a Heat Exchanger where it is cooled. Thereafter, the cooled gas10 is passed to a Spray Tower. The Spray Tower condenses the ammonia andnitrogen dioxide in an aqueous solution 11 which is sent to the AmmoniumHydroxide Purification Station, for further processing.

A 50% solution of sodium hydroxide 8 is diluted in a Mixing Tank withdeionized water 7 from the Deionized Water Supply 5. The resultingsolution 9 having a pH of 10-13.5 is introduced into a CO₂ ScrubberTower where it is used to scrub the carbon dioxide from the gas stream12 emerging from the Spray Tower, according to the following equation:

2NaOH+CO₂Na₂CO₃+H₂O

The sodium carbonate product solution 14 discharged from the CO₂Scrubber Tower is sent to Waste Treatment for disposal.

The remaining hydrogen-inert gas (nitrogen in the preferred route)mixture 13 emerging from the CO₂ Scrubber Tower is filtered. A membranesystem is then used to separate the excess hydrogen 15 from theremaining hydrogen-argon gas mixture 13. This separation is accomplishedwith relative ease due to the large molecular weight differences betweenthe hydrogen and argon gases and the selectivity of the membranes;hydrogen diffuses rapidly through membranes, and the other gases do notdiffuse at all. Typical membranes include polysulfone and related,hollow fiber designs. The excess hydrogen 15 flows to the Utility SupplyBuilding where it can be used as fuel in the incinerator. The purifiedhydrogen-argon mixture 16 is enriched with additional argon 20 from theArgon Supply and passed through the Heat Exchanger. The heated andenriched hydrogen-argon mixture 17 is divided into two portions 18 and19 that are sent to the VTF's in the 1,200 and 300 Micron Conversion andSintering Stations, respectively. Steam is added to reform (destroy)tramp oil into hydrogen, methane, and carbon. The flow sheet assumes aclosed system, but includes a 10% per year makeup for the argon.

Facility waste 21 from the VTF Gas Purification Station is sent to WasteTreatment.

Q. Ammonium Hydroxide Reconstitution

FIG. 4.20 depicts the flow sheet for the Ammonium HydroxideReconstitution Station. The material and energy balances for theexemplary process for production of uranium dioxide microspheres anduranium carbide microspheres are shown in Table 4.20 and Table 4.20a,respectively.

TABLE 4.20 Stream Number, Metric Tons per Year (Te/yr) Component 1 2 3 45 6 NH₄OH 18,578 2,797 0 21,375 14,717 6,658 Water 167,202 2,797 22,370192,369 132,451 59,918 Total 185,780 5,594 22,370 213,744 147,168 66,576Temp., ° C. 40 25 25 25 25 25 Flow, l/hr 17,700 532 2,554 20,300 14,0006,000

TABLE 4.20a Stream Number, Metric Tons per Year (Te/yr) Component 1 2 34 5 6 NH₄OH 18,578 2,797 0 21,375 14,717 6,658 Water 167,202 2,79722,370 192,369 132,451 59,918 Total 185,780 5,594 22,370 213,744 147,16866,576 Temp., ° C. 40 25 25 25 25 25 Flow, l/hr 17,700 532 2,554 20,30014,000 6,000

The ammonium hydroxide solution 1 from the Ammonium HydroxidePurification Station is reconstituted by mixing in a Mixing Tank with afresh, 50% ammonium hydroxide solution 2 and deionized water 3. Theresulting 10% solution of ammonium hydroxide 4 is divided into twofractions 5 and 6 that are sent to the 1,200 Micron and 300 MicronSetting/Washing Stations, respectively.

R. Urea and HMTA Recovery

FIG. 4.21 depicts the flow sheet for the Urca and HMTA Recovery Station.The material and energy balances for the exemplary process forproduction of uranium dioxide microspheres and uranium carbidemicrospheres are shown in Table 4.21 and Table 4.21 a, respectively.

TABLE 4.21 Stream Number, Metric Tons per Year (Te/yr) Component 1 2 3 45 6 Urea 5,596 280 5,316 14 5,316 0 HMTA 15,715 15,715 760 760 0 0 Water33,006 22,844 2,103 1,571 10,162 0 Facility Waste 0 0 0 0 0 2,2880 Total54,317 38,559 8,179 2,345 15,478 2,880 Temp., ° C. 40 40 40 40 40 25Flow, l/hr 5,167 3,650 557 223 1,472 2,880

TABLE 4.21a Stream Number, Metric Tons per Year (Te/yr) Component 1 2 34 5 6 Urea 5,596 280 5,316 14 5,316 0 HMTA 15,715 15,715 760 760 0 0Water 33,006 22,844 2,103 1,571 10,162 0 Facility Waste 0 0 0 0 0 2,2880Total 54,317 38,559 8,179 2,345 15,478 2,880 Temp., ° C. 40 40 40 40 4025 Flow, l/hr 5,167 3,650 557 223 1,472 2,880

Since urea and IIMTA effectively function as complexing agents andcatalysts the preferred embodiment, they are present in sizablequantities. Therefore, it is desirable to separate and recycle them. Inparticular, it is desirable to have a urea product that is relativelyfree of HMTA in order to avoid premature precipitation of uraniumdioxide. It is noted, however, that the recovered HMTA can contain ureawithout significant effects upon the overall process.

It is well known that urea and HMTA form crystals from concentratedsolutions. Thus, this property is often used in their manufacture andpurification. Urea possesses a solubility of approximately 50% at 17° C.in water, and 17% at 20° C. in alcohol. For HMTA, the solubility isabout 45% at 15° C. in water and 3% at 15° C. in alcohol. However, ureapossesses better crystallization properties, including the capability toform clathrate-type crystals (essentially double compound crystals) inthe presence of low concentrations of paraffinic hydrocarbons. Thus, inthe preferred embodiment, urea is selectively crystallized away from theHMTA, via a combination of thermal and chemical effects, and both arerecycled. Alternatively, tailored membranes may be effective for theurea-HMTA separation. The crystallization process may be advantageouslyoperated in the batch mode for enhancement of the separation.

S. Cylinder Decontamination

FIG. 4.22 depicts the flow sheet for the Urea and HMTA Recovery Station.The material and energy balances for the exemplary process forproduction of uranium dioxide microspheres and uranium carbidemicrospheres are shown in Table 4.22 and Table 4.22a, respectively.

TABLE 4.22 Section 1 Stream Number, Metric Tons per Year (Te/yr)Component 1 2 3 4 5 6 Cylinders 2,240 0 0 0 0 0 UO₂F₂ 52 0 0 0 0 52Water 0 8,960 8,960 100 Trace 8,960 HNO₃ 0 0 0 100 0 0 Facility Waste 00 0 0 0 0 UO₂(NO₃)₂ 0 0 0 0 0 0 Total 52 8,960 8,960 200 Trace 9,012Temp., ° C. 25 25 25 25 25 25 Flow, l/hr 1,023 1,023 1,023 15 Trace1,023

TABLE 4.22 Section 2 Stream Number, Metric Tons per Year (Te/yr)Component 7 8 9 10 11 Cylinders 0 2,18 112 168 0 UO₂F₂ 0 0 0 0 0 Water100 0 0 0 0 HNO₃ 5 0 0 0 0 Facility Waste 0 0 0 0 2,880 UO₂(NO₃)₂ <1 0 00 0 Total 100 2,128 112 168 2,880 Temp., ° C. 25 25 25 25 25 Flow, l/hr12 (972) (510 2.4 2,880

TABLE 4.22a Section 1 Stream Number, Metric Tons per Year (Te/yr)Component 1 2 3 4 5 6 Cylinders 2,240 0 0 0 0 0 UO₂F₂ 52 0 0 0 0 52Water 0 8,960 8,960 100 Trace 8,960 HNO₃ 0 0 0 100 0 0 Facility Waste 00 0 0 0 0 UO₂(NO₃)₂ 0 0 0 0 0 0 Total 52 8,960 8,960 200 Trace 9,012Temp., ° C. 25 25 25 25 25 25 Flow, l/hr 1,023 1,023 1,023 15 Trace1,023

TABLE 4.22a Section 2 Stream Number, Metric Tons per Year (Te/yr)Component 7 8 9 10 11 Cylinders 0 2,18 112 168 0 UO₂F₂ 0 0 0 0 0 Water100 0 0 0 0 HNO₃ 5 0 0 0 0 Facility Waste 0 0 0 0 2,880 UO₂(NO₃)₂ <0 0 00 Total 106 2,128 112 168 2,880 Temp., ° C. 25 25 25 25 25 Flow, l/hr 12(972) (510 2.4 2,880

The cylinder decontamination station removes the residual uranium(“heel”) from the cylinder via a three step operation. In the firststep, the empty cylinders 1 are filled continuously rinsed withdeionized water 2 from the Deionized Water Supply 3 for approximatelyfour hours. Since most of the remaining uranium fluorides andoxyfluorides are dissolved in this rinsing step, low decontaminationfactors (DF's) of 10-20 (i.e., 90-95% removal) and production of thehighly soluble, uranyl fluoride is expected. The rinse water 6, whichbecomes a dilute uranyl fluoride solution, is periodically recycled tothe Uranyl Fluoride Formation Station as part of the quench waterrequirements. In the second step, the empty cylinders are filled andrinsed with a dilute (5%) nitric acid solution 4 for four hours. Thenitric acid solution 4 allows ion exchange recovery of the uranium in aside column, thus regenerating the acid and allowing its reuse. Thisprovides for very high DF's of 50-1,000. Finally, in the third step, theempty cylinders are filled and rinsed with deionized water 5 from theDeionized Water Supply 3 for approximately four hours, to removeresidual chemicals and traces of uranium. This provides an additional DFof 2-10.

Facility Waste 11 from the Cylinder Decontamination Station is sent toWaste Management.

T. Waste Management

FIG. 4.23 depicts the flow sheet for the Urea and HMTA Recovery Station.The material and energy balances for the exemplary process forproduction of uranium dioxide microspheres and uranium carbidemicrospheres are shown in Table 4.23 and Table 4.23a, respectively.

TABLE 4.23 Section 1 Stream Number, Metric Tons per Year (Te/yr)Component 1 2 3 4 5 6 7 Facility Waste, 50,000 0 0 25,000 25,000 8,000 0liter/yr Cement 0 40 0 0 0 0 0 Water 0 0 20 0 0 0 0 Drums 0 0 0 0 0 40250 Air/N₂ 0 0 0 0 0 0 0 DAW 0 0 0 0 0 0 0 Activated 0 0 0 0 0 0 0Carbon Ion Exchange 0 0 0 0 0 0 0 Resins HEPA's 0 0 0 0 0 0 0 Total50,000 40 20 25,000 25,000 8000 250 Temp., ° C. 25 25 25 25 25 25 25Flow, liter/hr 5.7 2 2 3 3 1 5.7

TABLE 4.23 Section 2 Stream Number, Metric Tons per Year (Te/yr)Component 8 9 10 11 12 13 14 Facility Waste, 50,000 0 0 0 0 176,200 0liter/yr Cement 40 0 0 0 0 40 0 Water 20 98,624 14,003 0 0 20 112,623Drums 250 0 0 0 0 880 0 Air/N₂ 0 0 0 24,000 12.6E6 0 0 DAW 0 0 0 0 0 0 0Activated 0 0 0 0 0 0 0 Carbon Ion Exchange 0 0 0 0 0 0 0 Resins HEPA's0 0 0 0 0 0 0 Total 60 98,624 14,003 24,000 12.6E6 176,200 112,623Temp., ° C. 25 25 25 25 25 25 25 Flow, liter/hr 5.7 11,260 1,600 2.3E61.2E9 20 12,840

TABLE 4.23 Section 3 Stream Number, Metric Tons per Year (Te/yr)Component 15 16 17 18 19 20 Facility 0 0 0 0 0 0 Waste Cement 0 0 0 0 00 Water 0 0 0 0 0 0 Drums 250 0 0 0 201 140 Air/N₂ 0 12.6E6 0 0 0 0 DAW0 0 0 4E6 40,200 0 Activated 0 0 20,000 0 0 0 Carbon Ion Exchange 0 0 00 0 28,000 Resins HEPA's 250 0 0 0 0 0 Total 250 12.6E6 20,000 4E640,200 28,000 Temp., ° C. 25 25 25 25 52 25 Flow, liter/hr 5.7 1.2E9 3460 5 3

TABLE 4.23a Section 1 Component/ Stream Number, Metric Tons per Year(Te/yr) Item 1 2 3 4 5 6 7 Facility Waste, 50,000 0 0 25,000 25,0008,000 0 liter/yr Cement 0 40 0 0 0 0 0 Water 0 0 20 0 0 0 0 Drums 0 0 00 0 40 250 Air/N₂ 0 0 0 0 0 0 0 DAW 0 0 0 0 0 0 0 Activated 0 0 0 0 0 00 Carbon Ion Exchange 0 0 0 0 0 0 0 Resins HBPA's 0 0 0 0 0 0 0 Total50,000 40 20 25,000 25,000 8000 250 Temp., ° C. 25 25 25 25 25 25 25Flow, liter/hr 5.7 2 2 3 3 1 5.7

TABLE 4.23a Section 2 Component/ Stream Number, Metric Tons per Year(Te/yr) Item 8 9 10 11 12 13 14 Facility Waste, 50,000 0 0 0 0 176,200 0liter/yr Cement 40 0 0 0 0 40 Q Water 20 105,844 14,003 0 0 20 119,887Drums 250 0 0 0 0 880 0 Air/N₂ 0 0 0 24,000 12.6E6 0 0 DAW 0 0 0 0 0 0 0Activated 0 0 0 0 0 0 0 Carbon Ion Exchange 0 0 0 0 0 0 0 Resins HEPA's0 0 0 0 0 0 0 Total 60 105,844 14,003 24,000 12.6E6 176,200 119,887Temp., ° C. 25 25 25 25 25 25 25 Flow, liter/hr 5.7 12,091 1,600 2.3E61.2E9 20 13,700

TABLE 4.23a Section 3 Component/ Stream Number, Metric Tons per Year(Te/yr) Item 15 16 17 18 19 20 Facility 0 0 0 0 0 0 Waste Cement 0 0 0 00 0 Water 0 0 0 0 0 0 Drums 250 0 0 0 201 140 Air/N₂ 0 12.6E6 0 0 0 0DAW 0 0 0 4E6 40,200 0 Activated 0 0 20,000 0 0 0 Carbon Ion Exchange 00 0 0 0 28,000 Resins HEPA's 250 0 0 0 0 0 Total 250 12.6E6 20,000 4E640,200 28,000 Temp., ° C. 25 25 25 25 52 25 Flow. liter/hr 5.7 1.2E9 3460 5 3

Liquid waste streams 9 from the Oil Purification Station and the VTF GasPurification Station are carbon filtered to remove traces of oil andorganics The filtered waste stream 9 is then passed through an IonExchange to remove the traces of ionic species. Meanwhile, liquid wastestream 10 from the reverse osmosis is filtered. Permeate produced in theHMTA Recycle Station is similarly filtered. Waste streams 9 and 10 arethen merged into a single stream 14 which is essentially pure water, andare discharged to Publicly Owned Treatment Works (“POTW”) or NationalPollution Discharge Elimination System (“NPDES”) Point. No liquidradioactive wastes are generated by this process.

Solid wastes from the operations are collected, and sorted. Incinerationis used to treat combustible wastes, while non-combustible waste iscompacted and solidified.

MANUFACTURE OF DENSE URANIUM CARBIDE AND EXEMPLARY PROCESS

In the preferred embodiment, microspheres of depleted uranium carbidesare advantageously utilized as shielding materials, because they havethe highest densities of any uranium compounds. More particularly,uranium monocarbide (UC) is preferred, because it has slightly betterphysical properties than uranium dicarbide (UC₂), including a higherdensity and thermal conductivity. In either form, however, uraniumcarbides react slowly in moist air to form uranium dioxide. Thus, in thepreferred embodiment, it is desirable to apply an impervious coating tothe depleted uranium carbide microspheres that will render them inertunder normal conditions.

There are two principal routes for manufacturing uranium carbidematerials: (1) graphite reduction; and (2) gelation. Historically, theproduction of uranium carbide has been accomplished using the reductionof uranium dioxide with carbon (e.g., from graphite). This process isdescribed in M. Benedict, T. Pigford, and H. Levi, Nuclear ChemicalEngineering, Second Edition, McGraw-Hill, New York, N.Y., 1981,incorporated herein by reference. The uranium dioxide starting materialsutilized in the reduction process can be manufactured from uraniumhexafluoride or uranyl nitrate solutions using a variety of knownmethods. In the preferred embodiment, however, uranium carbide materialsare manufactured directly utilizing a gelation process that issubstantially similar to the uranium dioxide gelation process discussedpreviously.

The graphite reduction and gelation routes for production of uraniumcarbide are discussed in detail below.

A. Graphite Route

FIG. 5 prevents an overview of the graphite route for production ofdepleted uranium carbides.

As discussed above, the reduction of uranium dioxide to form uraniumcarbide is known in the art. In the preferred embodiment, a uraniumdioxide solid is mixed with carbon powder and an polyethylene binder toform a slurry. This slurry is then oven dried and ball-milled tosand-sized particles (0.03-2 mm). The oxides are then converted intocarbides in a vacuum heating operation, in which the oxygen is replacedby carbon, and, consequently, carbon monoxide and carbon dioxide arereleased. The amount of carbon in the initial mixture determines whetherthe uranium monocarbide or the dicarbide are formed. The resulting smallparticles of uranium carbide are fed through a furnace operating inexcess of the normal melting point of the carbide, and the microspheresare formed. Surface tension effects produce the spherical shape.

The depleted uranium dioxide starting materials used in the reductionprocess can be generated using any of a variety of known methods. In thepreferred embodiment, however, uranium dioxide is produced by gelation.While the gelation route for production of uranium dioxide can beadvantageously used to generate uranium carbide by reduction of uraniumdioxide microspheres, it is more desirable to produce the uraniumcarbides directly in the gelation process as described in detail below.

Once the depleted uranium carbide microspheres have been produced, it isdesirable to apply a coating which will effectively isolate the uraniumcarbide from the environment at the microscopic level. While multi-layercoatings are known in the art, the preferred embodiment utilizes asingle coating of carbon. This coating can be applied in a fluidized bedfurnace in which a stream of inert gas (usually argon) is introducedinto the furnace to levitate and heat the carbide microspheres. Amixture of hydrocarbons is then introduced into the gas stream. Thehydrocarbons dissociate when they come in contact with the outer surfaceof the microspheres, and form a dense, pyrolytic carbon layer.

B. Gelation Route

FIG. 4.1a depicts the modified flow diagram for the manufacture ofdepleted uranium carbide microspheres by gelation.

As is readily apparent from FIG. 4.1a, the process for production ofuranium carbide microspheres is substantially similar to the process forproduction of uranium dioxide shown in FIG. 4.1. There are four mainchanges to the gelation process when uranium carbide microspheres arethe desired end product: (1) a carbon powder/surfactant solution isprepared in a Carbon Suspension Formation Station (FIG. 4.6.1 and Table4.6.1); (2) the carbon powder/surfactant solution is added to the uranylnitrate solution in the Uranyl Nitrate Solution Adjustment Station (FIG.4.6.2 a and Table 4.6.2a) (3) sintering of the uranium carbide spheresin the Uranium Carbide Coating Station is accomplished using a two-stepsintering process (FIG. 4.24 and Table 4.24); and (4) the uraniumcarbide microspheres are coated in a Uranium Carbide Coating Station(FIG. 4.25 and Table 4.25).

TABLE 4.25 Stream Number, Metric Tons per Year (Te/yr) Component 1 2 3 45 6 UC₂ 20,840 0 0 0 24,040 0 Argon 0 1000 0 0 0 0 Methane 0 0 3,726 0 00 Propylene 0 0 0 4,141 0 0 Total 20,840 1000 3,726 4,141 24,040 0Temp., ° C. 25 25 25 25 25 25 Flow, 216 61,000 600,000 250,000 250 2,880liter/hr

As shown in FIG. 4.1a, depleted uranium hexafluoride is reacted withsteam to produce uranyl fluoride and hydrogen fluoride; the latter beingrecoverable in an anhydrous form. The solid uranyl fluoride iscollected, quenched, and dissolved in water. Adjustment of the residualhydrogen fluoride concentration is optionally undertaken utilizingdistillation methods. As with the uranium dioxide gelation process,uranyl fluoride can be utilized directly for the formation ofmicrospheres. Nevertheless, in a conservative approach, the uranylfluoride is converted to uranyl nitrate using calcium nitrate toprecipitate the fluoride and form uranyl nitrate in solution.

Meanwhile, a carbon suspension is formed in a Carbon SuspensionFormation Station. FIG. 4.6.1 depicts the flow sheet for the CarbonSuspension Formation Station. The material and energy balances for theexemplary process are shown in Table 4.6.1.

TABLE 4.6.1 Stream Number, Metric Tons per Year (Te/yr) Component 1 2 34 5 6 7 Carbon 2,850 0 0 0 2,850 0 0 Pigment Surfactant 0 14 0 0 14 0 0Water 0 0 11,386 640 (steam) 11,386 640 0 Facility Waste 0 0 0 0 0 0 0Total 2,850 14 11,386 640 14,250 640 0 Temp., ° C. 25 25 25 150 50 15025 Flow Rate, 81 2 1,300 73,000 1,356 73 0 Liters/hour

The manufacture of uranium carbide microspheres requires the addition ofcarbon to convert the uranium dioxide produced by the gelation processinto uranium carbide. In the preferred embodiment, carbon is introducedto the uranium solutions, prior to gelation, as part of an aqueoussuspension of fine carbon particles. As shown below, the carbonsubstitutes for the oxygen in the uranium dioxide and generates carbondioxide:

UO₂+2C→UC+CO₂

UO₂+3C→UC₂+CO₂

The required quantity of carbon is estimated from the stoichiometry ofthe conversion and sintering reactions; the monocarbide requires lesscarbon than the dicarbide. For the 100% capacity case, 1,900 tonnes/yearwould be required for the monocarbide and 2,850 tonnes/year for thedicarbide. The particulate form of carbon, such as carbon black or finegraphite, is preferred because of its small size, consistency, and easeof making a suspension. The quantity of added carbon depends upon thedesired final carbide (monocarbide or dicarbide), and is usually between14% and 25% of the broth solution (weight basis). Consequently,surfactants are added to the solution to stabilize the particles andkeep them in suspension. The carbon/surfactant suspension is added tothe uranium nitrate solution in the Uranyl Nitrate Adjustment Station,as shown in FIG. 4.6.2 a and Table 4.6.2a.

In the preferred embodiment, a 20% carbon suspension is prepared bycombining appropriate amounts of carbon pigment 1 and deionized water 3in a Mixing Tank. About 1000 ppm of surfactant 2 is added to the MixingTank to facilitate dispersion of the fine carbon particles in the water.Gentle heating of the suspension to about 50° C. using a Heat Exchangeralso assists the dispersion. The resulting suspension 5 is pumped to theUranyl Nitrate Solution Adjustment Station. FIG. 4.6.2 a depicts theflow sheet for the modified Uranyl Nitrate Solution Adjustment Stationfor uranium carbide production. The material and energy balances for theexemplary process are shown in Table 4.6.2a.

Adjustment of the uranyl nitrate solution for production of uraniumcarbide is virtually identical to the uranium dioxide process describedpreviously. In the modified process, however, the carbon/surfactantsuspension 14 from the Carbon Suspension Formation Station is introducedinto the VR Evaporator along with the uranyl nitrate solution 1 and theurea solution 3. The resulting solution is processed in the same maimeras before.

As with the uranium dioxide routes, the carbon-containing, adjusteduranyl nitrate solution is used for gel sphere formation. In thepreferred embodiment, internal gelation routes are preferred. Followinggel sphere formation, the gel spheres are aged in an ammonium hydroxidesolution. After aging, the gel spheres are dried at similar temperaturesto remove water and excess ammonia.

Subsequently, a vertical tube furnace (“VTF”) sinters the microspheresunder an argon-hydrogen atmosphere. In the preferred embodiment, a twozone furnace with an inert gas, such as nitrogen, is utiltized in orderto avoid over-reduction of the uranium. FIG. 4.24 depicts the flow sheetfor the modified Uranium Carbide Sintering Station. The material andenergy balances for the exemplary process are shown in Table 4.24.

TABLE 4.24 Component/ Stream Number, Metric tons per year (Te/yr) Item 12 3 4 5 (NH₄)₂U₂O₇ 24,540 0 0 0 0 UC₂ 0 0 20,840 0 0 NH₃ 0 0 0 1,930 0CO₂ 0 0 0 8,600 0 N₂ 0 0 0 trace 0 Argon 0 15,400 0 15,336 0 Hydrogen 0160 0 613 0 NO₂ 0 0 0 6.6 0 Urea 300 0 0 0 0 HMTA 830 0 0 0 0 NH₄OH 46 00 0 0 Oil 1,100 0 0 0 0 Nitrogen 0 0 0 0 0 Carbon 2,850 0 0 0 0 Water490 1,600 0 0 0 Facility Waste 0 0 0 0 0 Total 40,430 17,200 20,84026,500 0 Temp, ° C. 25 150 100 300 25 Flow, 1/hr 2,565 1,67E6 216 3.6E60

The sintering process for uranium carbide spheres is substantially thesame as uranium dioxide sintering, except that a Two-Zone VTF isutilized to avoid over reduction of the uranium carbide. Too muchhydrogen causes conversion the uranium carbide into uranium metal andmethane. The first zone of the VTF utilizes an argon cover gas 2containing 2-4% hydrogen. During this stage of the sintering process,most of the reaction and generation of carbon monoxide and dioxideoccur. It should be noted that sintering of uranium carbide spheresproduces substantially more carbon monoxide and carbon dioxide than withuranium dioxide. The second zone of the VTF operates at highertemperatures, using only argon as the cover gas, and results insintering and the high densities desired. The final sintered sphereshave densities usually exceeding 95% of the theoretical density foruranium carbides. If two or three sizes of microspheres are produced,then spatial densities exceeding 90% of theoretical can be obtained byvibratory loading methods.

Subsequently, as with the graphite route, fluidized bed furnaces applycoatings to the microspheres which effectively isolate the uraniumcarbide from the environment at the microscopic level. While the massand energy balances in Table 4.24 are set forth for two coatings, it isexpected that a single coating will be preferred.

PEROXIDE GELATION

Peroxide gelation is an alternative gelation process contemplated inaccordance with the present invention. FIG. 6 provides an overview ofthe peroxide gelation process. As shown in FIG. 6, uranium hexafluorideis vaporized and defluorinated to produce anhydrous hydrogen fluorideand uranyl fluoride powder. However, the steam required for the reactioncomes from a recycle stream 1 containing the azeotrope (HF.2H₂O) plustraces or uranyl fluoride, nitric acid, and aluminium nitrate. Theuranyl fluoride powder is quenched and dissolved in diluted, aqueousnitric acid and used to dissolve uranium metal and low density oxidefeed materials. Dissolution is aided by a nitric acid recycle stream 2.Fluorboric acid and urea may be added to the solution in thequench/dissolving step. Aluminum nitrate can be added in the molar ratioof 0.001 to 1.25 to facilitate partial complexation of the fluorideions. The stream is chilled to 0 to 25° C., and dispersed using nozzlesinto a hydrogen peroxide solution bath or column. The peroxide solutionhas a concentration between 0.5 to 50%, and is maintained between 0-45°C. Uranyl peroxide (UO₄.2H₂O) precipitates as a microsphere. The solidsare separated by screens and filters washed with a dilute peroxidestream (0.001 to 5 M), dried in warm nitrogen and sintering undernitrogen to produce dense uranium dioxide microspheres.

Uranium carbides can be manufactured by adding a carbon suspension tothe uranium stream prior to chilling and droplet formation. The uranylperoxide precipitate particle retains the carbon pigment, but allows thesoluble species such as nitrates, fluorides, and urea to diffuse intothe bulk solution. The two-step sintering procedure, describedpreviously, is used to produce the dense carbides, but no hydrogen isneeded.

The peroxide solution recycles between the precipitator and thefilter/separator. The recycled solution contains hydrogen fluoride,nitric acid, aluminum nitrate, fluorboric acid, and traces of uranylfluoride and peroxide. Once the peroxide is consumed by reaction orotherwise depleted by decomposition the stream is periodically orcontinuously withdrawn and distilled. The bottoms product fromdistillation contains the hydrogen fluoride-water azeotroph (boilingpoint circa 110° C.), with aluminium nitrate, fluorboric acid, andtraces of uranyl fluoride/nitrates. This stream is recycled to thedefluorinator. The distillate product from distillation contains nitricacid, at temperatures of 50 to 75° C. Molecular sieves, distillation, orrelated means can be used to remove excess water prior to recycle of thenitric acid stream 2 to the quencher.

Although a particular form of the invention has been illustrated anddescribed, it will be appreciated by those of ordinary skill in the artthat various modifications, alterations, and substitutions can be madewithout departing from the spirit and scope of the invention.Accordingly the scope of the present invention is not to be limited bythe particular embodiments set forth above, but is only to be defined bythe following claims.

We claim:
 1. A shielding material precursor, comprising: a particulateuranium compound selected from the group consisting of uranium dioxideparticles and uranium carbide particles and mixtures of uranium dioxideparticles and uranium carbide particles; and a thermosettingpolyfunctional resinous binding material.
 2. The precursor of claim 1,wherein the uranium compound comprises at least 5 weight % of theprecursor.
 3. The precursor of claim 1, wherein the uranium compoundcomprises about 55 weight % to 80 weight % of the precursor.
 4. Theprecursor of claim 1, wherein the resinous binding material is apolyimide.
 5. The precursor of claim 1, wherein the resinous bindingmaterial is a polyamide.
 6. A shielding material precursor, comprising:a particulate uranium compound selected from the group consisting ofuranium dioxide particles and uranium carbide particles and mixtures ofuranium dioxide particles and uranium carbide particles; and a resinousbinding material, wherein said resinous binding material ispolyfunctional.
 7. The precursor of claim 6, wherein the resin is apolyimide.
 8. The precursor of claim 6, wherein the resin is apolyamide.
 9. The precursor of claim 6 wherein the resin ispolyurethane.
 10. A shielding material precursor, comprising: aparticulate uranium compound selected from the group consisting ofuranium dioxide particles and uranium carbide particles and mixtures ofuranium dioxide particles and uranium carbide particles; and a metalbinding material.
 11. The precursor of claim 10, wherein the metalbinding material is selected from the group consisting of copper, zinc,nickel, tin, aluminium, boron, and mixtures thereof.
 12. A shieldingmaterial precursor, comprising: a particulate uranium compound selectedfrom the group consisting of uranium dioxide particles and uraniumcarbide particles and mixtures of uranium dioxide particles and uraniumcarbide particles; and a metal-oxide binding material.
 13. The precursorof claim 8, wherein the metal-oxide is selected from the groupconsisting of alumina, boric acid, magnesia, silica, hafnia, hematite,magnetite, and zirconia.
 14. The precursor of claims 1, 6, 10, or 12wherein the uranium compound is coated.
 15. The precursor of claim 14wherein the uranium compound is coated with carbon.
 16. The precursor ofclaims 1, 6, 10, or 12, wherein the precursor further comprises ashielding additive.
 17. The precursor of claim 16, wherein the shieldingadditive comprises up to about 20 weight % of the precursor.
 18. Theprecursor of claim 16, wherein the shielding additive is selected fromthe group consisting of hydrogen, boric acid, sodium borate, gadoliniumoxide, halfnium oxide, erbium oxide, and indium oxide.
 19. The precursorof claim 16, wherein the shielding additive is steel shot.
 20. Theprecursor of claim 16, wherein the shielding additive is glass beads.21. The precursor of claims 1, 6, 10, or 12, wherein the uraniumcompound particles are microspheres formed by gelation.
 22. Theprecursor of claims 1, 6, 10, or 12, wherein the uranium compoundparticles have at diameters from about 30 to 2000 microns.
 23. Theprecursor of claims 1, 6, 10, or 12, wherein the diameters of theuranium compound particles are within one of at least two discreteranges of particle sizes.
 24. The precursor of claim 23, wherein thefirst particle size range is from about 300 to 500 microns and thesecond particle size range is from about 1000 to 1300 microns.
 25. Theprecursor of claims 1, 6, or 10, wherein the uranium compound comprisesa mixture of uranium dicarbide and uranium dioxide particles.
 26. Theprecursor of claim 25, wherein the uranium carbide and uranium dioxideparticles are coated.
 27. A precursor of claim 25, wherein the uraniumdioxide particles are coated.
 28. A precursor of claim 27, wherein theuranium carbide particles are coated with carbon.
 29. A precursor ofclaim 27, wherein the uranium dioxide particles are coated with carbon.30. The precursor of claim 25, wherein the uranium carbide comprises upto 70 weight % of the uranium compound mixture.
 31. A precursor of claim30, wherein the uranium carbide comprises particles having a size rangewherein the diameter of the particles is from about 1000 to 1300 micronsand wherein the uranium dioxide comprises particles have a size rangewherein the diameter of the particles is from about 300 to 500 microns.32. A method for forming a monolithic shielding material, comprising:combining a particulate uranium compound, a metal oxide binder, andwater to form a mixture; curing the mixture at a sufficient temperatureand pressure to form a monolithic shielding material.
 33. The method ofclaim 32, wherein water comprises up to about 40 weight % of themixture.
 34. The method of claim 32, wherein the mixture is cured atambient temperature and pressure.
 35. The method of claim 32, whereinthe mixture is cured by heating the mixture.
 36. The method of claim 35,wherein the mixture is cured by heating the mixture at a temperaturefrom about 100° C. to about 400° C.
 37. The method of claims 32 or 35,wherein curing the mixture further includes applying pressure to themixture.
 38. The method of claim 32, wherein curing the mixture is curedat a pressure up to about 20 atmospheres.
 39. The method of claim 32,wherein curing the mixture further includes providing a combustiblematerial in proximity to the mixture and heating the mixture whereby thecombustible wall is consumed.
 40. A method for forming a monolithicshielding material, comprising: combining a particulate uranium compoundselected from the group consisting of uranium dioxide particles anduranium carbide particles and mixtures of uranium dioxide particles anduranium carbide particles and a metal binder to form a precursormixture; and curing the mixture at a sufficient temperature and pressureto form a monolithic shielding material.
 41. The method of claim 40,wherein the mixture is cured by heating.
 42. The method of claim 41,wherein the mixture is cured by heating the mixture at a temperaturefrom about 400 to 1000° C.
 43. The method of claim 41, wherein themixture is cured by induction heating.
 44. The method of claims 40 or41, wherein curing the mixture further includes applying pressure to themixture.
 45. The method of claim 40, wherein the mixture is cured at apressure up to about 20 atmospheres.
 46. A method for forming amonolithic shielding material, comprising melting a metal bindingmaterial; adding a particulate uranium compound selected from the groupconsisting of uranium dioxide particles and uranium carbide particlesand mixtures of uranium dioxide particles and uranium carbide particlesto the melted binder to form a mixture; cooling the mixture to form amonolithic shielding material.
 47. The method of claim 46, wherein themetal binding material is selected from the group consisting of copper,zinc, nickel, boron and mixtures thereof.
 48. A method for forming amonolithic shielding material, comprising combining a particulateuranium compound particulate uranium compound selected from the groupconsisting of uranium dioxide particles and uranium carbide particlesand mixtures of uranium dioxide particles and uranium carbide particlesand a thermosetting polyfunctional resin binder to form a precursormixture; and curing the mixture at a sufficient temperature and pressureto form a monolithic shielding material at ambient temperature andpressure.
 49. The method of claim 48, wherein the mixture is cured byheating.
 50. The method of claim 49, wherein the mixture is cured byheating the mixture at a temperature from about 400 to 600° C.
 51. Themethod of claim 49, wherein the mixture is cured by induction heating.52. The method of claims 48, 49, or 50, wherein curing the mixturefurther includes applying pressure to the mixture.
 53. The method ofclaim 52, wherein the mixture is cured by applying a pressure up toabout 20 atmospheres.
 54. A method for forming a monolithic shieldingmaterial, comprising: combining a particulate uranium compound selectedfrom the group consisting of uranium dioxide particles and uraniumcarbide particles and mixtures of uranium dioxide particles and uraniumcarbide particles and a polyfunctional resin binder to from a precursormixture; and curing the mixture at a sufficient temperature and pressureto form a monolithic shielding material.
 55. The method of claim 54,wherein the mixture is cured by heating.
 56. The method of claim 55,wherein the mixture is cured by heating the mixture at a temperaturefrom about 400 to 600° C.
 57. The method of claim 55, wherein themixture is cured by induction heating.
 58. The method of claims 54, 55,56 or 57, wherein curing the mixture further includes applying pressureto the mixture.
 59. The method of claims 54, 55, 56, or 57, whereincuring the mixture further includes applying a pressure up to about 20atmospheres.
 60. A monolithic shielding material, comprising: apyrolized uranium compound selected from the group consisting of uraniumdioxide particles and uranium carbide particles and mixtures of uraniumdioxide particles and uranium carbide particles, and a thermosettingpolyfunctional resinous binding material.
 61. A monolithic shieldingmaterial, comprising: a pyrolized uranium compound and a metal bindingmaterial, said uranium compound being selected from the group consistingof uranium dioxide particles and uranium carbide particles and mixturesof uranium dioxide particles and uranium carbide particles.
 62. Amonolithic shielding material, comprising a pyrolized uranium compoundand a metal oxide binding material.
 63. A shielding material precursorcomprising: a particulate uranium compound selected from the groupconsisting of uranium dioxide particles and uranium carbide particlesand mixtures of uranium dioxide particles and uranium carbide particles;and a binding material selected from the group consisting of athermosetting polyfunctional resinous binding material, metal bindingmaterial, and metal oxide binding material; and wherein the bindingmaterial constitutes up to 95 weight % of the precursor material. 64.The shielding material precursor of claim 63 wherein the bindingmaterial constitutes 20-45 weight % of the precursor and the particulateuranium compound constitutes 55-80 weight % of the precursor material.